本企画調査では、放射線、金属材料、熱流動、電気化学などが複合した全く新しい現象である放射線誘起表面活性の研究の方向性を確認するその一歩として、各専門研究者を集約した共同研究体制により、1)原子炉内放射線誘起熱伝達技術調査・分析、2)放射線誘起熱伝達メカニズム調査分析、3)放射線誘材料の調査・分析、4)宇宙環境における放射線誘起表面活性の調査・分析、5)放射線誘起腐食防止効果の技術調査・分析、及び6)放射線誘起電気特性調査・分析が行われ、文献1-12に示されるように、非常に大きな成果を得た。
その成果をもとに、本企画調査の主要な目的であった科学研究費補助金平成16年度発足『特定領域研究』申請「放射線誘起表面活性による機能創出技術の構築」がなされている。この特定領域研究の目的は、放射線・電磁波・材料・電気化学・腐食・伝熱流動・計測などの研究分野の研究者を有機的に結合する研究領域を設定することにより、放射線誘起活性現象のメカニズム解明及び最適化を行い、放射線誘起表面活性による新たな利用分野の創成特に非放射線環境における放射線規制値未満の微弱放射能による自励放射線誘起表面活性によって生成される機能、創出技術を構築することである。この特定領域研究の実施により、(1)発展段階の観点から見て成長期の初期段階にあり、学術研究における先導的又は基盤的意義を有する研究領域である放射線誘起表面活性の利用研究の急速な展開、及び(2)放射線誘起表面活性現象は多くの研究分野が関係しているために、この領域の研究の発展は互いの研究分野に新しい情報を提供し、各学術研究分野の水準向上・強化につながり、かつそれらの発展に大きく寄与すると、それぞれ考えられる。また放射線誘起表面活性技術をエネルギー環境技術として確立することにより、我が国が技術立国として今後も存在し、世界に貢献できる重要な手段が確立されると期待される。

研究代表者らは、酸化ジルコニウムなどの広いバンドギャップを有する被膜にγ線を照射することにより、励起電子が被膜に接する金属に移行して電位を卑化させる放射線誘起表面活性現象を見いだした。本研究の目的は、この現象を利用して、金属の腐食を緩和する方法を実験的に確認することである。
実験に用いられた試験片形状は全て幅20mm、長さ50mm、厚さ1mmであり、母材は純度99.99%の鉄、SUS304L、およびSUS316Lである。これに酸化被膜として酸化チタン、酸化ジルコニウム、または酸化アルミニウムを表面に溶射により成膜した。表面のみを試験対象とするため、裏面および端部にはエポキシ系樹脂を塗布した。3wt%塩化ナトリウム水溶液中に17時間浸漬した酸化アルミニウムを約200μm溶射した99.99%鉄試験片は、暗室保管の場合、一部孔食が見られ、ほぼ全面に腐食が進行した。一方、γ線を照射した場合には、このような腐食様相はほとんど見受けられなかった。溶液浸漬時間を40h,64hとした実験を行った結果、暗室の場合には腐食が更に進行したが、γ線照射の場合は腐食の進行が遅いことが判明した。これらの結果から、放射線照射により顕著な防食効果を発揮することが示された。線量率を増大させることにより、一層高い腐食緩和効果が生じ、また酸化チタン以外の酸化ジルコニウムおよび酸化アルミニウムについても、γ線照射により高い防食特性が得られることがわかった。
このように炉内構造物表面の酸化金属被膜(SUS、ジルカロイなど)がγ線と反応して、放射線誘起表面活性を生じ、腐食の進行を低減できることが確認された。このような技術のさらなる開発は特に、原子炉高経年化使用、近年社会的問題になった原子炉構造物のひび割れ防止、また原子炉安全上大きな課題となっている応力腐食割れの抑制を目的に開発を急ぐ必要がある。

本研究の目的は,液滴の加熱面衝突時において遷移沸騰域で認められる多数の微小な液滴が大気圧雰囲気中へ激しく飛散する微細化現象を明らかにすることにある.特に液滴と加熱面間に生成された薄い蒸気膜が崩壊して固液接触を起こし,そこで発生する微小気泡(マイクロバブル)の発生挙動をとらえ,微細化の機構と関連づけようとするものである.これまでに使用した石英加熱面では核沸騰的な挙動が高温域まで続き,激しい微細滴の飛散現象が顕著に認められなかったため,サファイア円板,ガラス面に薄く導電性皮膜を施したITO(Indium Tin Oxide:酸化インジウム スズ)皮膜面及び石英面に薄くクロム皮膜を施した面を加熱面として使用し,それぞれの面上での沸騰挙動を各裏面から高速度カメラでとらえ,石英面と比較した.
1.ITO皮膜面及びクロム皮膜面での核沸騰域(加熱面と液膜とが十分に濡れた状態下の沸騰)の蒸発寿命特性は石英面でのデータと類似している.これに対して熱物性値が石英の4.9倍に相当するサファイア面では加熱面温度範囲は低温側に移行し,金属面での寿命特性に接近している.
2.ITO皮膜面上での沸騰挙動に関して,核沸騰域では,液膜内に微小な蒸気泡の発生とみなされる白い斑点が多数観察された.これらの白い点は液膜の中央部から放射状に周辺部へ速い速度で移動しているのが判明した.
3.サファイア面上での沸騰挙動に関しは,加熱面温度が320℃以下では,激しい核沸騰的な比較的大きい蒸気泡の発生が認められたが,370℃以上では薄い蒸気膜が生成されその上に液滴が存在すると言った状況が確認でき,液膜の表面から放射状に微小な液粒子の速い飛散が認められた.
4.クロム皮膜面では450℃を越えると,分裂した液滴の一部が加熱面と薄い蒸気膜を介して接触を保つようになり,その蒸気膜の崩壊時に爆発的な微小液粒子の飛散が生ずることが確認できた.これを誘起するものはマイクロバブルであり,これは薄い蒸気膜崩壊に伴う固液接触時に生ずることが判明した.

© 2019 Elsevier Ltd Void fraction (i.e., the volume fraction occupied by gas) is a key parameter for determining the coolability and neutron-moderating performance of a water-cooled nuclear reactor. To develop computational multi-fluid dynamics models for determining the void-fraction distribution, experimental data of comparable quality are required. We have developed a high-energy X-ray computed tomography (CT) system to acquire three-dimensional void-fraction distributions. The CT system comprises a linear-accelerator-driven high-energy X-ray source and a linear detector array. We quantified a boiling two-phase flow in a 5 × 5 heated rod bundle at high pressure, simulating a fuel-rod bundle in a boiling water reactor (BWR). Because the axial travel of the CT system is 4 m and includes the entire BWR fuel-rod bundle, we optimized the CT imaging conditions and reconstruction method for rod-bundle visualization to reduce uncertainties due to density fluctuations in the boiling flow and imaging artifacts. We conducted a boiling experiment at a low flow rate and low thermal power and acquired three-dimensional distributions of the void fraction over a wide pressure range of 0.1–7.2 MPa. The experiment provided three-dimensional void-fraction distributions with high spatial resolution, especially in subchannel regions surrounded by rods, and the results are suitable for validating three-dimensional thermal-hydraulic analysis codes.

When boiling occurs in a pool or gas is injected into a stagnant water pool, the actual liquid level, i.e., the two-phase mixture level, becomes higher than the original liquid level, i.e., the collapsed water level, and depends on the gas held in the liquid pool. The two-phase mixture level is an important indicator of the core cooling of a nuclear reactor and of the long-term operation of the filtered containment venting system under accident conditions. To clarify the relationship between the two-phase mixture and collapsed water levels, we conducted an air-water two-phase flow experiment in which we used different channel geometries, i.e., circular pipes and rod bundles, to inject air at atmospheric pressure into stagnant water. Then, we used visual observation to obtain the swell and fluctuation amplitude of the two-phase mixture level. The results indicate that with increases in the superficial gas velocity, the flow channel geometry significantly affects the swell of the two-phase mixture level, and that the dominant bubble scale in the rod bundle was influenced by both the hydraulic diameter and the channel-box scale. We compared the height and fluctuation amplitude of the two-phase mixture level with existing and newly acquired experimental data and models.

In a postulated sever accident of a light water reactor (LWR), molten core debris (corium) may breach the reactor pressure vessel and be released to the ex-vessel containment floor. A core catcher manages ex-vessel corium cooling by uniformly spreading the corium into a large space, but it requires a dedicated plant design. In contrast, corium shields have been back-fitted to some boiling water reactors to prevent excessive amount of corium to flow into sump pits, where effective corium cooling may be difficult. However, corium shields can only block the corium flow and cannot contribute to uniform spreading of the ex-vessel corium. This study proposes a preliminary concept of “Debris Spreading Floor”, which can be applied to any types of reactor plants including back-fitting to the existing plants. More specifically, the existing containment floor is overlaid with sacrificial material and refractory material is placed around sump pits. It is intended to allow the original function of sump pits to collect leaking water under normal, abnormal transient and design basis accident conditions. However, under postulated severe accident condition, spreading of ex-vessel corium is promoted by ablating itself with the hot corium and guiding corium spreading away from sump pits. To develop the concept, mechanistic analysis of corium spreading, which can consider influence of substrate ablation, is needed. The Moving Particle Semi-implicit (MPS) method is a Lagrangian particle method and thus suitable for mechanistic simulation of free-surface spreading flow involving solid / liquid phase change and interactions. In this research, effect of the proposed concept is firstly presented with the MPS simulations. Preliminary simulations in 2D show that, amount of corium flowing into sump pits is reduced by the concept. Secondly, validity of the MPS simulations is quantitatively discussed by simulating the experiments with simulant. The experiment was carried out by Central Research Institute of Electric Power Industry (CRIEPI) by pouring liquid Pb-Bi onto a Pb-Bi block so that the inflow liquid spreads on the block surface while it also ablates the block. Sensitivity analyses have been carried out with different initial conditions, calculation resolutions, subscale models and parameters of the MPS simulations to identify the key models and parameters for quantitative prediction of the melt / substrate interactions.

Numerical analyses were conducted to replicate several tests for simulating a double-ended cold-leg large break loss-of-coolant accident (LBLOCA) in the Loss-of-Fluid Test (LOFT) using the TRACE (version 5/patch level 4) code. Analytical results by the original TRACE code were so conservative that especially a first peak of cladding temperature was estimated higher than the experimental data at the blowdown phase and subsequent temperature drop corresponding to the temporal quench was not seen. We were interested in minimum film boiling temperature (T min ) as a heat transfer model factor estimating the quench at the moment, investigated correlation equations for T min in previous studies and especially focused on ones given as a function of coolant mass flow because the complicated flow transient and decompression in the core region at the blowdown phase was interpreted as having an influence on the cladding temperature behavior. There are several correlations meeting the above condition but it was revealed that they are insufficient to apply for high pressure especially. Therefore, a new term including an effect of mass flow flux and time derivative of pressure was defined and added with a proportional coefficient hypothetically to the current correlation in the TRACE code for modification. The LOFT analyses were conducted again using the modified TRACE code, and it was shown by applying roughly the same proportional coefficient to all the cases of LOFT analyses that estimation of the cladding temperature behavior was improved more precisely at the blowdown phase. Also, the transition during the phase was explained phenomenologically with the wall heat transfer mode and boiling curve.

<p>Highly accurate measurement of thermal properties, reaction rates and mass transfers is necessary to grasp thermal flow characteristics of nuclear reactor plant because these properties are unsteadily sensitive to concentration, temperature, pressure and scale. On the real scale of the plant, however, there are problems such as the complexity of three-dimensional flow and the effect of convection. In this study, a novel technology that measures mass transfer and reaction rate of nuclear application material with high accuracy and high speed by applying microchannel which makes one-dimensional laminar flow, has developed. In this paper, a micro flow control system that can adjust flow with a minimum volume of 1 μL/min. by a pressurized pump, was fabricated. In order to observe concentration diffusion of the flow field in real time, a near infrared spectroscopic visible observation device was developed and combined with the flow system. The mutual diffusion coefficient was measured for hydrazine hydrate, which is added for the purpose of preventing corrosion of secondary piping in nuclear power plant. It revealed that the absorption depends on the concentration near 1555 nm for Fourier transform near-infrared spectrophotometer. The diffusion coefficients of 5, 10 and 20 wt% for hydrazine hydrate were quantified with the luminance profiles in the absorption peak wavelength.</p>

© 2018 Elsevier B.V. Rapid thermal elevation in boiling water reactor (BWR) is an important factor for nuclear safety and there is a need to develop an analysis code for the transient phenomenon and its validation process. To evaluate the thermal property of transient boiling and its uncertainty, corroborative experimental information is crucial. In particular, the lateral propagation behavior of a vapor bubble (void) in the cross-sectional direction of fuel assembly has yet to be determined. This study evaluates the void propagation behavior in a 5 × 5 rod bundle with cross-sectional heat distribution that causes only the 3 × 3 rod bundle to generate heat; assuming rapid heating under atmospheric pressure. In this paper, using the maximum heat output applied to the nine heated rods as a parameter, from the visualization of the void behavior and the measurement of the local void fraction, the heat output conditions under circumstances where lateral propagation of voids occurs and where voids are only localized in the heated region are summarized. We quantified the time difference initially detected and the time-averaged void fraction according to the lateral propagation level.

In order to investigate heat transfer deterioration due to salt precipitation, a pool boiling experiment was conducted with a full-height 5×5 rod-bundle of BWR simulated fuel in sea water. The temperature on the center rod surface at the top spacer rose rapidly, since the flow area inside the top spacer was filled with the precipitated salt. Dryout occurred below the spacer, which results in the temperature escalation of the heater surface. On the other hand, the heater above the top spacer was cooled continuously by pool boiling.

It is necessary to simulate a eutectic melting reaction and relocation behavior of boron carbide (B₄C) as a control rod material and stainless steel (SS) during a core disruptive accident in an advanced sodium-cooled fast reactor designed in Japan. On that account, a new project has been started to conduct eutectic melting experiments, thermophysical property measurement of the eutectic melt, and physical model development for the eutectic melting reaction. The eutectic experiments involve the visualization experiments, eutectic reaction rate experiments and material analyses. The thermophysical properties are measured in the range of liquid and solid states. The physical model is developed for a severe accident computer code based on the measured data of the eutectic reaction rate and the physical properties. This paper describes the project overview and progress by JFY2017.

<p>A pedestal-floor structure of a nuclear-reactor containment-vessel was additive manufactured with two sump-pits and a doorway at a reduced scale of 1:100 and 1:50. A silicone-oil jet impinged and spread on the pedestal floor. Such three-dimensional flow was visualized from three cameras. A volume of fluid (VOF) simulation was successfully reproduced this three-dimensional spreading behavior, silicone-oil flow-thickness evolution and drainage weights.</p>

This chapter addresses characteristics of flashing-induced density wave oscillations on the basis of the experimental results in a boiling natural circulation system with an adiabatic chimney. Flashing is caused by the sudden increase of vapor generation due to the reduction in hydrostatic head, since saturation enthalpy changes with pressure. Flashing-induced density wave oscillations may, therefore, occur at low pressure. The oscillation period correlates well with the passing time of bubbles in the chimney section regardless of the system pressure, the heat flux, and the inlet subcooling. According to the stability map, the flow became stable below a certain heat flux regardless of the channel inlet subcooling. The stable region enlarged with increasing system pressure. Therefore, the stability margin becomes larger by pressurizing the loop sufficiently before heating.

Accident analyses of unit 3 in the Fukushima-Daiichi nuclear power station were performed with MAAP (Modular Accident Analysis Program) 5.03. In the analysis, the operations of RCIC, HPCI, SRV were simulated and assuming the operation of alternative water injection to reproduce the measured pressure and temperature. As a result of parametric evaluation, analysis results consistent with measured values are obtained. In this paper, the results of the sensitivity analysis are reported. Simultaneously, attempts were made to analyze the transient and deposition amount of fission product (FP) in the reactor building (R/B) and a model dividing each room of R/B in detail was created to confirm the FP behavior. Based on these analyses, the deposition amount of the primary containment vessel (PCV) (drywell (D/W) and wetwell (W/W)) and R/B could be estimated using detailed model of the R/B dividing into nodes in the MAAP simulation.

Copyright © 2018 ASME. Rapid thermal elevation in nuclear reactor is an important factor for nuclear safety. It is indispensable to develop a three-dimensional nuclear thermal transient analysis code and confirm its validity in order to accurately evaluate the effectiveness of the running nuclear safety measures when heating power of reactor core rapidly rises. However, the heat transfer characteristics such as reactivity feedback characteristics due to moderator density and the technical knowledge explaining the uncertainty are insufficient. In particular, the cross propagation behavior of vapor bubble (void) in cross section of fuel assembly is not grasped. This study evaluates the cross propagation void behavior in a simulated fuel assembly at time of rapid heat generation with a thermal hydraulic test loop including a 5×5 rod bundle having the heat generation profile in the flow cross sectional direction. In this paper, the branching heat output condition of transient cross propagation was investigated from visualization of high speed video camera and void fraction measurement by wire mesh sensor with the inlet flow rate 0.3m/s and the inlet coolant temperature 40oC, which are based on the transient safety analysis condition. In addition, we applied the particle imaging velocimetry (PIV) technique to measure liquid-phase velocity profile of the coolant in the transient cross flow and experimentally clarified the relationship with the cross flow.

Accident analyses of the Fukushima-Daiichi unit-2 nuclear power plant were performed with MAAP (Modular Accident Analysis Program) version 5.03. We assumed RCIC, SRV operation and alternative water injection in order to reproduce the measured pressure and temperature values in RPV and PCV. From parametric studies, it was found that the analysis results were in good agreement with the measured data. In this paper, the results of the parametric studies are reported. Furthermore, spatial discretization of compartments (such as rooms in the reactor building, etc.) into small parts successfully demonstrated the transient distribution and deposition of fission products (FPs) across the rooms. Such special discretization is particularly important for the forensic investigation of severe accidents and the deposited amount in the R/B might be estimated by using this detailed model.

When light water reactor (LWR) is subject to a cold shutdown, it needs to be cooled with pure water or seawater to prevent the core melting. To precisely evaluate the cooling characteristics in the fuel assembly, a measurement method capable of installing to the fuel assembly structure and determining the temperature distribution with high temporal resolution, high spatial resolution, and in multidimension is required. Furthermore, it is more practical if applicable to a pressure range up to the rated pressure 16 MPa of a pressurized water reactor (PWR). In this study, we applied the principle of the wire-mesh sensor technology used in the void fraction measurement to the temperature measurement and developed a simulated fuel assembly (bundle) test loop with installing the temperature profile sensors. To investigate the measurement performance in the bundle test section, it was confirmed that a predetermined temperature calibration line with respect to time-average impedance was calculated and became a function of temperature. To evaluate the followability of measurement in a transient temperature change process, we fabricated a 16 × 16 wire-mesh sensor device and measured the hot-water jet-mixing process into the cold-water pool in real time and calculated the temperature profile from the temperature calibration line obtained in advance from each measurement point. In addition, the sensors applied to three-dimensional temperature distribution measurement of a complex flow field in the bundle structure. The axial and cross-sectional profiles of temperature were quantified in the forced flow field with nonboiling when the 5×5 bundle was heated by energization.

Vapor explosion has been causing disasters in many industries such as metalwork and paper industries. One of the countermeasures is retardant additives into water to stabilize the vapor film which separates two liquids. A spontaneous vapor explosion of a molten tin jet at 700 oC was suppressed with only 0.03 wt% polyethylene glycol aqueous solution for molecular weight of 4×106 g/mol. This is because the solute deposited near the vapor-liquid interface due to the cloud-point phenomenon, that stabilizes vapor film and prevents the solution from mixing finely. Salts are known additives to act as vapor-explosion promoter. Increasing salt concentration requires denser PEG solution to suppress vapor explosion: e.g. 0.03 wt% PEG for water, while 0.07 wt% PEG for sea water and 3 wt% sodium chloride aqueous solution. These salt solutions were selected for practical relevance in industrial disasters. A solid sphere quenching experiment indicates that this threshold concentration of PEG can be determined by the quenching temperature of the solid sphere: the contact temperature of the solid sphere with solution must be sufficiently low (e.g. spontaneous-bubble nucleation temperature of the solution) to suppress the vapor explosion.

This chapter deals with the various topics on boiling with regard to aspects of the fundamentals and applications to introduce the development of each author's research in recent decades. The first four sections investigate the physics of boiling as phase change phenomena, including thermodynamic phase equilibrium state (Section 6.1), molecular dynamics of phase change (Section 6.2), computational analysis of boiling in micro-nano scale (Section 6.3), and transient boiling under rapid heating (Section 6.4). Section 6.5 deals with two-phase distribution measurement using neuron radiography. The following three sections then examine a specific boiling regime during highly subcooled boiling, called microbubble emission boiling (MEB). Each section treats the overall characteristics of MEB (Section 6.6), the occurrence conditions of MEB (Section 6.7), and vapor collapses in subcooled liquid related to MEB (Section 6.8). The next four sections are devoted to heat transfer augmentation with various techniques: thermal spray coating (Section 6.9), porous media (Section 6.10), patterned wettability refinement (Section 6.11), and self-rewetting fluid (Section 6.12). The last seven sections describe topics on applications of boiling. Sections 6.13 and 6.14 introduce boiling research in steel industries. Sections 6.15 and 6.16 explore vapor explosion. Boiling of refrigerant is discussed with heat pump systems in Section 6.17 and with automobile air conditioners in Section 6.18. Boiling related to emergency cooling core systems is considered in Section 6.19.

The fuel support piece in a boiling water reactor (BWR) is used to brace fuel assemblies. The channel within the fuel support piece is determined to be a potential corium relocation path from the core region to the lower head during the severe accident of BWR. In the present study, the improved ∗∗∗Moving Particle Semi-implicit (MPS) method was adopted to simulate the flow and solidification behavior of the melt in a fuel support piece. The MPS method was first validated against the Pb-Bi plate ablation test that was performed by CRIEPI. The predicted ablation mass of the plate agreed well with the experimental results. Then the flowing and freezing behaviors of molten stainless steel (SS) and zircaloy in the fuel support piece were simulated by MPS method with a three dimensional particle configuration, respectively. In this study, the flow and solidification behavior of SS was simulated first. After all the SS passed through the channel, the flowing behavior of Zr in the fuel support piece was simulated. The simulation results indicated that the crust layer formed on the inner surface of the fuel support piece during the melt discharging process. The fuel support piece was plugged by the solidified zircaloy particles in the lower initial temperature case. The fuel support piece kept intact in all the calculation that were performed under the assumed order of melt injection. The present results could help to reveal the progression of a BWR severe accident.

<p>A filtered containment venting system (FCVS) reduces containment pressure and amount of the radioactive release to the environment during a severe accident of nuclear power plant. CRIEPI has developed a FCVS analysis tool which evaluate removal performance of aerosol, iodine and organic iodine in each process of FCVS as decontamination factor as a function of the FCVS system parameters. The paper addresses the integrated analysis tool of FCVS simulator based on the experimental database.</p>

The dose assessment of the main control room has been required in order to show its residence under the present Japanese new regulatory standards. Furthermore, to assess the in-plant dose such as reactor building where a lot of operations are performed is considered to contribute to the evaluation of accessibility in the training scenario for future voluntary safety improvements. The severe accident analysis code MAAP (Modular Accident Analysis Program) has "DOSE" module that dose calculation and accident progress analysis can be implemented simultaneously. It is expected to contribute in the evaluation of more detailed accident scenario because temporal trend of the dose corresponding to FP (Fission Product) amount can be obtained for each accident sequence. This study intends to obtain the knowledge to handle the DOSE module of MAAP by performing a dose calculation of the reactor building of a BWR plant. The FP migration behavior and the dose rate in the reactor building and the main control room were evaluated against two accident sequences, the only SBO and the superimposition of SBO and LOCA, by using BWR4 Mark-I containment model which is distributed as a sample of MAAP. The dose calculation result reflecting the accident progress was obtained from the result such as the LOCA and SBO sequence showed relatively high doses because of the early pressure vessel damage. The interphase between the dose and the FP mass was good for each accident sequence, especially for the noble gas. The SBO sequence showed the overall higher dose and the FP amount flowing into the reactor building was also larger. Noble gases, CsOH and CsI were confirmed as FP nuclides with a high degree of influence on dose, and the contribution of noble gas with large release amount was particularly higher. Since the noble gas is released to the external environment along with the time, the contribution of CsOH and CsI became higher in the latter stage of the scenario. In addition, as one of the factors that have an impact on the dose rate in the building during a severe accident, drywell spray operation is considered. The effect of spray on dose can be considered as below, the capture of FP, cooling of the debris, and the shielding effect of the water. The effect of greatly reducing the dose in the SBO scenario was confirmed. In addition, it was confirmed that the cooling debris by spray had a large influence on the dose.

In order to quantify thermo-physical flow field for the industrial applications such as nuclear and chemical reactors, high temporal and spatial measurements for temperature, pressure, phase velocity, viscosity and so on are required to validate computational fluid dynamics (CFD) and subchannel analyses. The paper proposes a novel temperature profile sensor, which can acquire temperature distribution in water at high temporal (a millisecond) and spatial (millimeter) resolutions. The devised sensor acquires electric conductance between transmitter and receiver wires, which is a function of temperature. The sensor comprise wire mesh structure for multipoint and simultaneous temperature measurement in water, which indicated that three-dimensional temperature distribution can be detected in flexible resolutions. For the demonstration of the principle, temperature profile in water was estimated according to predetermined temperature calibration line against time-averaged impedance. The 16×16 grid sensor visualized fast and multi-dimensional mixing process of a hot water jet into a cold water pool.

Fuel claddings would be exposed to a hightemperature steam-flow after a boiled-up sequence of a severe accident of light water reactors as well as spent fuel pools. Zirconium alloys are common cladding material for light water reactors. All the zirconium alloys generate additional heat by the exothermic reaction of oxidation in high-temperature steam. Subsequent temperature rise accelerates the rate of oxidation. The countermeasure to suppress oxidation of zirconium alloy is key to mitigate severe accidents. We invent a surface modification technology of zirconium alloy, Fresh Green, to oxidize and to carbonize a zirconium-alloy surface in the same process. XRD and XPS analyses indicated that a modified layer on zirconium surface is carbon-doped zirconium dioxide, in which some of oxygen atoms in monoclinic zirconium dioxide are replaced by carbon. Experiments are conducted for three representative conditions using an autoclave at relatively high temperature: uniform corrosion at 400 degrees Celsius for 336 hours, nodular corrosion at 500 degrees Celsius for 24 hours, supercritical water corrosion (at 400 degrees Celsius, 24 hours). The Fresh Green surface modification reduces both the oxidation rate and hydrogen-pickup less than a half of that with untreated base material. This is because the Fresh Green layer is closely packed and adhered intimately to the base material. Hydrides were found in the base material without treatment, though they were scarcely observed in the Fresh Green treated specimen. Since the operating temperature of the Fresh Green process is lower than the final annealing temperature in a manufacturing process, the process does not affect the material property and crystal structure. The Fresh Green process, which can be operated at slightly higher pressure than ambient, can be introduced in the conventional manufacturing process without complicity.

The effectiveness evaluation of the safety measures under severe accidents is demanded; not only for nuclear reactors but also spent fuel pools (SFPs) after the Fukushima-Daiichi nuclear power plant accident which hit the Tokyo Electric Power Company on March 11, 2011. This paper addresses models of the spent fuel pool with accident analysis code, MAAP and a sensitivity analysis of the spray cooling parameters following a loss of coolant accident (LOCA) to evaluate the fuel coolability. In this study, spray water is injected from the upper space of SFP onto fuel racks using an SFP model attached to a MAAP code. The instantaneous LOCA in SFP was hypothetically assumed, which meant the spent fuels were directly exposed to the atmosphere from the onset of the event. Furthermore, the sensitivity of the spray cooling parameters was analyzed to investigate the influence on how the instantaneous LOCA sequence in the SFP will progress. The spray cooling model of the MAAP code includes a range of input parameters such as the spray-flow rate, spray water temperature, and so on. For example, when the spray-flow rate suffices to cool the fuel, the maximum temperature of the fuel cladding decreases after the onset of spray cooling with a particular time delay. Furthermore, increasing the spray-flow rate helps reduce any time delay and accelerate cooling. Conversely, when the spray-flow rate is insufficient for cooling, the fuel cladding temperature continues to rise and ends up damaging the fuel. The spray-flow rate significantly impacts on the cooling of the fuel accordingly. The MAAP code demonstrates a time trace of the maximum cladding temperature in terms of spray cooling parameters. In addition, certain spray cooling parameters may impact on the behavior of fission products (FP). For example, particles drifting in the atmosphere in aerosol form can be easily adsorbed by spray droplets when the spray-flow rate increases, which further eliminates FP. We confirmed the impact on FP removal of the spray.

Steam explosion has been a potential threat to containment boundary during severe accident in light water reactors. Polyethylene glycol (PEG) aqueous solution is an effective triggering retardant for both spontaneous and external-triggered steam explosions. In order to investigate steam explosion controllability at high temperature and for high meting-point metal, experiments are extended for molten nickel and tin up to 1800oC. Steam explosion was suppressed for both molten tin and nickel in a 0.1 wt% PEG (molecular weight of 4×106 g/mol) aqueous solution. The stability of PEG was investigated with gel permeation chromatography, ion chromatography, and high-performance liquid chromatography. The results indicated that PEG may be precipitated in the liquid side without producing byproducts and suppress steam explosions even for the molten jet at 1800 oC.

It is expected that filtered containment venting system (FCVS) would play an important role to reduce the amount of fission products (FPs) released to the environment during severe accident conditions. Experimental studies for FCVS have been conducted in CRIEPI, and decontamination factors (DFs) regarding internal impaction at nozzles, Venturi’s scrubbers, metal fibrous filters, and Zeolite filters have been measured and formulated, respectively. The DF models were implemented into the MELCOR input data assuming a typical advanced boiling water reactor (ABWR). Several accident progression analyses, such as large break loss-of-coolant accident (LOCA), long term station blackout (LTSBO), and loss-of-all power supply, were conducted for the BWR with a FCVS that consists of scrubbers, metal and Zeolite filters connected to its suppression pool and dry well. The result indicated that as the water level of the FCVS was decreased, efficiency of scrubber was decreased, and the total DF was governed by metal fibrous filter. After most of the water in FCVS was evaporated, FCVS temperature increased, and FPs once captured by the FCVS and deposited on wall inside of the system were transported the environment. Time-depended release ratios of Cs, CsI, Te, Sr to the environment were compared under condition with FCVS and that without FCVS. The results obtained effectiveness of FCVS for each FP. The results also gave some insights for implementation of FCVS and severe accident management for BWRs.

In order to design boilers, evaporators and condensers, a phase distribution is one of the most important factor to be determined. Although electrodes which are immersed into a coolant media give useful information such as void fraction and phasic velocity, one may suffer temperature variation which affect electric conductance of the coolant. Potential gradient and electromagnetic wave which are generated in a test liquid due to energization, give critical measuring accuracy in the form of electric noise. If the test liquid has temperature dependency of electric conductivity, time-displacement of liquid temperature during the transient boiling phenomenon gives additional uncertainty to the measured value. Figure 1 illustrates experimental apparatus, which consists of a pre-heater, test section, separator, condenser and circulation pump. A heater pipe is inserted in the test section and heated by direct current. Wire-Mesh Sensors (WMSs) were inserted near by the heated section. The experiment was conducted in non-boiling (no bubbles in a solution) condition to quantify temperature dependence of each solute. The test fluids are dilute aqueous solutions, whose solutes are vitriolic solutes (K2SO4 and Na2SO4), nitric solutes (KNO3 and NaNO3) and fluorescence (Rhodamine-B and Uranine) which is used in PIV, LDV. The experimental procedures are tabulated as follows. 1. Calibration using ion-exchanged water - rest potential of water was measured at intervals of 10° from 30° to 90°. The value became larger in high temperature condition. - 2. Setup of solution concentration (Table 1) - Gain value of detector was set 1, and the concentration of each solute was calibrated that the electric potential of its solution is equal to one of water in the case of 90°. - 3. Measurement of temperature dependency of electric potential (dV/dT, V: Voltage, T: Temperature) and SN ratio 4. Measurement of electric conductivity, pH and dissolved oxygen level (not mention in this extended abstract) Results and discussion were given as follows. Time-series data of electric potentials of water and Rhodamine-B solution were indicated in Figure2. As shown in these trends, these all data were normalized with time-averaged one of 90°. It was confirmed that amplification method of electric potential with solute concentration predominates that one method with detector gain in order to improve SN ration of the signal. Temperature sensitivity index (dV/dT) and SN ratio were illustrated in Figure 3. It shows the normalized dV/dT data when one of water is 100. It was clarified that the solutes except Uranine have the temperature dependency-reduction effect to 70 % of one water. Furthermore, SN ratio of all solutions were increased about 3 times than one of water. Finally, thermos-physical properties of test liquid solutions were illustrated in Figure 4 and 5, respectively. These graphs were normalized with the values at 30° each other. Fig. 4 indicates that addition of solute gives reduction effect of temperature change of electric conductivity. Especially in the case of Rhodamine-B, as shown in Fig. 5, its temperature dependence was indicated equal tendency with one of water because this fluorescence is not ionized. As the measurement of void fraction, reducing temperature dependency (dV/dT) without changing thermos-physical properties of water is important in order to avoid unnecessarily mixing the sensitive properties of electric conductivity. In conclusion, it was identified that Rhodamine-B has the following effects, reduction of temperature dependency of electric potential, improvement of SN ratio without changing water properties (pH, DO).

During a severe nuclear power plant accident, the release of fission products into containment and an increase in containment pressure are assumed to be possible. When the containment is damaged by excess pressure or temperature, radioactive materials are released. Pressure suppression pools, containment spray systems and a filtered containment venting system (FCVS) reduce containment pressure and reduce the radioactive release into the environment. These devices remove radioactive materials via various mechanisms. Pressure suppression pools remove radioactive materials by pool scrubbing. Spray systems remove radioactive materials by droplet-aerosol interaction. FCVS, which is installed in the exhaust system, comprises multi-scrubbers (venturi-scrubber, pool scrubbing, static mixer, metal-fiber filter and molecular sieve). For the particulate radioactive materials, its size affects the removal performance and a number of studies have been performed on the removal effect of radioactive materials. This study has developed a new means of evaluating aerosol removal efficiency. The aerosol number density of each effective diameter (light scattering equivalent diameter) is measured using an optical method, while the decontamination factor (DF) of each effective diameter is evaluated by the inlet outlet number density ratio. While the applicable scope is limited to several conditions (geometry of test section: inner diameter 500 mm × height 8.0 m, nozzle shape and air-water ambient pressure conditions), this study has developed a numerical model which defines aerosol DF as a function of aerosol diameter (d) and submergences (x).

<p>During a severe nuclear power plant accident, increasing of containment pressure are assumed to be possible. When the containment is damaged by excess pressure or temperature, radioactive materials are released from the containment. A filtered containment venting system (FCVS) reduces containment pressure and the radioactive release into the environment. The FCVS removes radioactive materials via various process. A pool scrubbing is an important process and flow dynamics (inertial collision, submergence, bubble diameter) and chemical conditions (pH, solubility) affects the radioactive materials removal performance. For an iodine (I2), chemical conditions are important factor in this process. This study has evaluated the relationships between the iodine decontamination performance and the chemical conditions in the pool scrubbing process.</p>

<p>In order to investigate boiling heat transfer of a porous body with a crack, alumina porous body was additivemanufactured on the basis of powder bed fusion method. Alumina was selected for its heat resistant property and similarity of a crust of molten core. The manufactured porous body with 30% porosity has a slit-shape crack. Additive manufacturing became possible with the optimized operating parameters, powder size distribution, and the specific sintering assistant. A metal foil was attached on the back side of porous body and heated by applied direct electric current. The temperature profile of the metal foil was estimated on the basis of acquired electric-resistance distribution. As a results of pool boiling experiment in a salt-water pool, the boiling heat-transfer performance was acquired as a function of heat-transfer levels.</p>

<p>During a severe nuclear power plant accident, increasing of containment pressure are assumed to be possible. When the containment is damaged by excess pressure or temperature, radioactive materials are released from the containment. A filtered containment venting system (FCVS) reduces containment pressure and the radioactive release into the environment. The FCVS removes radioactive materials via various process. In the pool scrubbing process, flow dynamics (inertial impact, submergence, bubble diameter) and chemical conditions (pH, solubility) affect the removal performance of the radioactive materials. This study has forcued on the aerosol removal effect by an inertial impact.</p>

A subchannel void sensor (SCVS) was improved to acquire time-resolve boiling two-phase flow in a 5-by-5 heated rod bundle at high-pressure and high-temperature. The SCVS consists of 6-by-6 wire electrodes and 5-by-5 rod electrodes. The SCVS can acquire a time series data of local void fraction at 32 points of central subchannel region in addition to 100 points of rod surface regions between rods. The devised sensors are installed in the heated rod bundle at eight height levels to acquire two-phase flow dynamics and its development. The axial and radial power profile of the heated rod bundle are uniform, and eight pairs of sheath thermocouples are embedded on the heated rod to monitor its surface temperature distribution. The paper addresses the axial and cross-sectional distributions of void fraction at various pressure and rod bundle power conditions, which simulates an accidental condition in reactor core and in a spent fuel pool of a boiling water reactor (BWR). The experimental data are suitable to validate the computational multi-fluid dynamics (CMFD) and subchannel codes.

Polyethylene glycol (PEG) was found to be a reliable retardant of steam explosions. In order to investigate the effects of molecular weight of PEG, concentration and salt additives on the controllability of steam explosions, experiments were conducted for tin drops immersed in a solution pool. Steam explosion was suppressed with a 0.03 wt% PEG solution for molecular weight of 4 million. This is because the cloudy-point phenomenon stabilizes vapor film and prevents the solution from mixing finely by the precipitated solute near the steam-water interface. The molecular weight must be selected in reference to the cloudy-point temperature to be lower than saturation temperature by a certain degrees at the target pressure. Steam explosion may occur in a PEG solution by adding 1 wt% of sodium chloride, because such salts act as steam explosion promoter and reduce the cloudy-point temperature significantly.

VOF numerical simulations were conducted for a three-fluid dam break problem to investigate the scaling effect on mixing and stratification processes of three immiscible fluids by gravity. Two liquids (silicone oil and salt water) were initially separated with a vertical partition plate and filled in a rectangle container. Computational fluid dynamics analysis was conducted with Star CCM+ version 9.02. The contact angle affects the interfacial motion significantly in a relatively small-scale domain, while the interfacial tension affects it in a relatively large-scale domain.

We devised the Subchannel void sensor (SCVS), which consists of 6-by-6 wire electrodes and 5-by-5 rod electrodes to acquire boiling two-phase flow dynamics in a 5-by-5 rod bundle at high-pressure and high-temperature conditions. SCVS can acquire a time series data of local void fraction at 32 points of central subchannel region and at 100 points of near rod surface regions. The experimental results exhibit the axial and cross-sectional distributions of void fraction, which are suitable to validate the CFD and subchannel codes.

In order to investigate ablation failure mode of lower head with molten core during a severe accident of light water reactors, ablation and melting relocation experiments were conducted with a hemisphere vessel with a drain hole. Three different silicone oil were used to investigate an effect of fluid viscosity to simulate a molten core. The hemisphere vessel is molded with lead bismuth eutectic alloy. The vessel wall thinning and melt relocation occurred just below the silicone oil level by ablation due to natural convection. For the lowerviscosity silicone oil, it results in breaking all around the vessel wall. On the contrary for the higher-viscosity silicone oil, the drain hole were ablated as well which enhance drainage flow. The time series of ablated molten wall and silicone oil weights drained from the hole were quantified separately on the basis of measured volume and weight of drained fluids for the code validation.

In the event of a nuclear power plant accident pressure within the containment can increase. A Filtered Containment Venting System (FCVS) allows for over-pressure release through multi-scrubbers (Venturi-scrubber, bubbling-scrubber, metal fiber filter and molecular sieve) and reduces the radioactive release. However, FCVS performance changes depending on operational conditions, e.g. steam flow rate, pressure, operating time and so on. The Central Research Institute of Electric Power Industry (CRIEPI)full-height FCVS test facility is constructed to measure FCVS performance under several conditions and can evaluate the decontamination factor (DF) of three major targets (aerosol, elemental iodine (I_2) and organic iodine (CH_3I)). This project is intended to acquire a systematic database of FCVS performance and optimize the FCVS operation procedure. The CRIEPI test vessel is about 8 m high, with an internal diameter of 0.5 m. FCVS performance tests were conducted under the following conditions: maximum pressure and temperature of 0.8 MPa and 170℃, inlet gas flow of steam (〜1600 kg/h) and air (〜300 kg/h) and containing aerosol/ iodine/ organic iodine. If fission product iodine gas is released into the environment during a severe accident, it will have a major impact on public health. This paper addresses the iodine decontamination performance by the bubbling effect. Iodine is effectively soluble in an alkaline solution. Accordingly, 0.5 wt% sodium hydroxide (NaOH) or a mixture of 0.2 wt% sodium thiosulfate (Na_2S_2O_3) and 0.5 wt% NaOH is used as an iodine filter (absorber) and during the experiment, an alkaline solution with composition equivalent to the actual equipment is used. The concentration of elemental iodine is quantified with an Inductively-Coupled Plasma with Mass Spectrometry (ICP-MS), while iodine DF is defined by the concentration ratio at the inlet and outlet. Iodine DF shows low dependence on flow dynamics, but dependence on solution property. Where iodine concentration is low, DF is high (between 10^4 and 10^5) and vice versa when the iodine concentration (saturate) increases.

In accidents when the water level of the reactor core descends below the top of the active fuel, the cooling limit height is a key factor in determining the accident mitigation procedure and the boiling two-phase flow in a fuel-rod bundle exhibits multi-dimensional and complex flow structures during such boil-off process. A rod bundle boil-off experiment was conducted to determine the three-dimensional void-fraction distribution and axial profile of the rod-surface temperature during the boil-off process under atmospheric pressure conditions. The 5×5 rod bundle, featuring a heated length of 2 m, had an axially and radially uniform power profile, with eight pairs of sheath thermocouples embedded on the heated rod to monitor its surface temperature distribution. The void-fraction distribution was acquired with five pairs of SubChannel Void Sensor (SCVS) as time-series data. The experimental results showed the relationship between an effective cooling level and boiling two-phase flow dynamics in the rod bundle.

In order to investigate ablation failure mode of lower head with molten core during a severe accident of light water reactors, ablation and melting relocation experiments were conducted with a hemisphere vessel with a drain hole. Three different silicone oil were used to investigate an effect of fluid viscosity to simulate a molten core. The hemisphere vessel is molded with lead bismuth eutectic alloy. The vessel wall thinning and melt relocation occurred just below the silicone oil level by ablation due to natural convection. For the lower-viscosity silicone oil, it results in breaking all around the vessel wall. On the contrary for the higher-viscosity silicone oil, the drain hole were ablated as well which enhance drainage flow. The time series of ablated molten wall and silicone oil weights drained from the hole were quantified separately on the basis of measured volume and weight of drained fluids for the code validation.

Copyright © 2015 by JSME. In accidents when the water level of the reactor core descends below the top of the active fuel, the cooling limit height is a key factor in determining the accident mitigation procedure and the boiling two-phase flow in a fuel-rod bundle exhibits multi-dimensional and complex flow structures during such boil-off process. A rod bundle boil-off experiment was conducted to determine the three-dimensional void-fraction distribution and axial profile of the rod-surface temperature during the boil-off process under atmospheric pressure conditions. The 5×5 rod bundle, featuring a heated length of 2 m, had an axially and radially uniform power profile, with eight pairs of sheath thermocouples embedded on the heated rod to monitor its surface temperature distribution. The void-fraction distribution was acquired with five pairs of SubChannel Void Sensor (SCVS) as time-series data. The experimental results showed the relationship between an effective cooling level and boiling two-phase flow dynamics in the rod bundle.

In the event of a nuclear power plant accident pressure within the containment can increase. A Filtered Containment Venting System (FCVS) allows for over-pressure release through multi-scrubbers (Venturi-scrubber, bubbling-scrubber, metal fiber filter and molecular sieve) and reduces the radioactive release. However, FCVS performance changes depending on operational conditions, e.g. steam flow rate, pressure, operating time and so on. The Central Research Institute of Electric Power Industry (CRIEPI) full-height FCVS test facility is constructed to measure FCVS performance under several conditions and can evaluate the decontamination factor (DF) of three major targets (aerosol, elemental iodine (I2) and organic iodine (CH3I)). This project is intended to acquire a systematic database of FCVS performance and optimize the FCVS operation procedure. The CRIEPI test vessel is about 8 m high, with an internal diameter of 0.5 m. FCVS performance tests were conducted under the following conditions: maximum pressure and temperature of 0.8 MPa and 170°C, inlet gas flow of steam (∼1600 kg/h) and air (∼300 kg/h) and containing aerosol/iodine/organic iodine. If fission product iodine gas is released into the environment during a severe accident, it will have a major impact on public health. This paper addresses the iodine decontamination performance by the bubbling effect. Iodine is effectively soluble in an alkaline solution. Accordingly, 0.5 wt% sodium hydroxide (NaOH) or a mixture of 0.2 wt% sodium thiosulfate (Na2S2O3) and 0.5 wt% NaOH is used as an iodine filter (absorber) and during the experiment, an alkaline solution with composition equivalent to the actual equipment is used. The concentration of elemental iodine is quantified with an Inductively-Coupled Plasma with Mass Spectrometry (ICP-MS), while iodine DF is defined by the concentration ratio at the inlet and outlet. Iodine DF shows low dependence on flow dynamics, but dependence on solution property. Where iodine concentration is low, DF is high (between 104 and 105) and vice versa when the iodine concentration (saturate) increases.

In the case of an accident and when a water level of a reactor core falls below the top level of active fuel, the cooling limit height becomes a key factor for determining the accident mitigation procedure. To predict the cooling limit height, it is important to clarify a two-phase mixture level in a rod bundle during the boil-off process. The two-phase mixture level depends on the collapsed level and void fraction distribution. During the boil-off process, a boiling two-phase flow in the rod bundle exhibits multidimensional and complex flow structures. The paper addresses a three-dimensional void fraction distribution and a two-phase mixture level in 5 × 5 heated rod bundles during the boil-off process, under atmospheric pressure conditions. The heated rod length is 3.7 m, which is the same as the fuel rod in boiling water-reactor (BWR). The 5x5 rod bundles have an axially and radially uniform power profile, and eight pairs of sheath thermocouples are embedded in the heated rod to monitor their axial surface temperature profiles. The diameter of the heated rod is 10 mm, and the rod pitch is 13 mm. The void fraction distributions were acquired with eight pairs of subchannel void sensors (SCVS) as time series data. The two-phase mixture level was evaluated by side-viewing images acquired with two high-speed digital video cameras. The experimental result exhibits a relationship of the boiling two-phase flow dynamics to the two-phase mixture level, and the void fraction during the boil-off process.

Steam explosion has been a potential threat during severe accident in light water reactors. We had reported that Polyethylene glycol (PEG) is a reliable retardant of steam explosions. Experiments were conducted to investigate the effects of concentration, molecular weight and salt additives on the controllability of steam explosions. Steam explosion was suppressed with a 0.03 wt% PEG solution for molecular weight of 4 million. This is because the cloudy-point phenomenon stabilizes vapor film and prevents the solution from mixing finely by the precipitated solute near the steam-water interface. The stabilizing effect of vapor film was confirmed in a solid stainless-steel sphere quenching experiment as well. The molecular weight must be selected in reference to the cloudy-point temperature to be lower than saturation temperature by a certain degrees at the target pressure. At atmospheric pressure, a molecular weight of 4 million is demonstrated to suppress steam explosions. The effective concentration became denser when large share stress and/or external force act on the vapor film. Steam explosion may occur in a PEG solution by adding lwt% of sodium chloride, because such salts act as steam explosion promoter.

In a severe accident of a light water reactor, ablation of the reactor pressure vessel (RPV) lower head by corium is a key phenomenon, which affects progression of the accident. The Moving Particle Semi- implicit (MPS) method is one of particle methods that calculate behavior of incompressible fluid by semi-implicit method. In preceding studies, MPS models have been developed to analyze phenomena such as heat conduction, phase change, natural convection, thermal stratification, and radiation heat transfer. These phenomena are expected to play key roles in the lower head ablation. This paper aims to investigate whether the MPS method is capable of analyzing the lower head ablation phenomenon, which involves complex interactions of the above mentioned phenomena. The small-scale experiment carried out at Central Research Institute of Electric Power Industry (CRIEPI) using Pb-Bi vessel and silicone oil was analyzed. The heat transfer model was modified for evaluation of heat transfer between the vessel and the oil. The results were compared both qualitatively and quantitatively with the experiment. The former is the comparison of the simulation and experiment regarding phenomena that the liquid ablates the metal vessel and discharges through the vessel wall, which showed good agreement. The latter are comparisons of the calculated liquid temperature, ablation start time and discharge start time with respect to the corresponding measurements. The analyses have shown that the MPS method is capable of analyzing ablation phenomenon qualitatively, but needs further development for quantitative prediction, including investigations on influence of the particle size used in the simulation.

© 2015 by Begell House, Inc. The boiling two-phase flow in the fuel rod bundle of a boiling water reactor exhibits multidimensional and transient flow dynamics. Coolability of the fuel is the key to ensuring safety, particularly in the event of any accident that involves the coolant injection malfunctioning and the core becoming uncovered. This paper addresses boiling two-phase flow dynamics in a 5 £ 5 heated rod bundle under atmospheric pressure. The diameter of the heated rod is 10 mm, the rod pitch is 13 mm, and the heated length is 3.71 m. The radial power profile is the key experimental parameter: (i) uniform, (ii) center peak, (iii) side peak, and (iv) corner peak. The cross-sectional void-fraction distribution of a total of 132 points was acquired at more than 800 frames (cross sections) per second with a SubChannel Void Sensor (SCVS). The axial void-fraction distribution was also acquired with eight pairs of SCVS, which explain the radial and axial development of boiling two-phase flow in the rod bundle.

SubChannel Void Sensor (SCVS) consisting of electrodes with 6-wire by 6-wire and 5-rod by 5-rod electrodes has been developed to acquire boiling two-phase flow dynamics in 5×5 rod bundle geometry. The SCVS can acquire a local void fraction in 32 points (=6×6-4) of central subchannel regions and 100 points (=4×25) of near rod surface regions. The temporal resolution is up to 5000 frames (cross sections) per second. The devised sensors are installed every 500mm in the heated rod bundle and exhibits the quasi three-dimensional boiling two-phase flow structures, i.e. void fraction distribution.

Two experimental campaigns were conducted to observe mixing and stratification processes of two immiscible fluids by gravity. Parametric study reviles the effects of two key fluid properties: kinematic viscosity (or molecular weight) for silicone oil and density (or concentration) for sodium chloride aqueous water. After withdrawal of a vertical partition plate, two liquids intersects earlier at the center, while the fluids stick on the walls. Another experiments were conducted with the salt water injecting on to the silicone oil pool. The kinematic viscosity and density difference affect three-dimensional mixing and stratification processes significantly. These visual databases are suitable for a code validation on the interfacial phenomena. Key Words : Gravity-driven stratification, Injection, Mixture, Phase separation, Liquid-liquid, Flow visualization

Severe accident analysis was performed for a station blackout (TBU) sequence in a typical BWR-5 plant with modified Mark-II type containment by MAAP5.01 code. Core heatup, degradation and relocation processes, reactor vessel failure and resultant impact on the containment including hydrogen production were investigated. Sensitivity analysis focusing on direct contact heating (DCH), which affect on the consequence of TBU sequence, was also performed. If extensive DCH does not occur in the TBU sequence, failure of the containment vessel can be postponed. On the other hand, concrete ablation at a floor and a side wall inside the pedestal due to molten core - concrete interaction (MCCI) significantly increases, because a large amount of debris with high temperature stays inside the pedestal.

The Moving Particle Semi-implicit (MPS) method is improved to analyze the stratification behavior of highly viscous fluids. The stratifying processes of silicone oil and salt water are simulated in 2D and 3D respectively, and the time history of fluid shapes is compared with experimental results. Characteristic fluid shapes for high and low silicone oil viscosity conditions are well predicted and the results show good agreement with experiments.

The carbon-doped copper oxide is devised as an electrocatalyst to reduce carbon dioxide under ambient pressure and temperature. The electrode was immersed into a KHCO₃ aqueous solution with CO₂ bubbling. The electric potential was maintained at -1.64 V vs SHE. The electrode prepared at 900&deg;C gives the maximum production rate of ethylene (25 %), ethanol (6.9 %) and 1-propanol (3.6 %). The production rate of methane, from which is harmful to separate ethylene, was suppressed to one fifteenth of that of ethylene. In contrast to a thermally-oxidized copper-oxide layer, the doped-carbon and a high ratio of Cu₂O to CuO in the devised electrocatalyst may result in the higher productivity and selectivity.

内径224mm大口径円筒管内二相流の軸方向発達過程を二枚セットのワイヤメッシュセンサにより計測した。得られた三次元気泡信号に対して、ワイヤメッシュセンサ間で類似の気泡を追跡する気泡追跡法を開発し、この手法により気泡径分布・三次元速度分布を評価した。

燃料の溶融移動挙動を把握するとともに解析の検証データを得るため、円柱状金属試験片を融点以上の雰囲気に設置し、溶融変形をビデオ観察した。変形の時系列変化を金属物性値と対比して考察した。

福島第一原子力発電所では、全電源喪失により燃料棒の冷却が出来なくなり炉心溶融に至った。溶融した燃料は圧力容器下部にデブリとして堆積していると考えられており、従って、廃炉措置を行うにあたり、溶融燃料取り出しや再臨界評価のためのデブリ分布等の情報が必要である。本研究では、福島第一原子力発電所廃炉措置等に資するため、燃料溶融及び溶融燃料の落下挙動に着目し、炉心構成要素である燃料棒・チャンネルボックス・制御棒の溶融過程を解析可能である粒子法をもとにした数値解析手法の開発を目的とする。本報では、本解析手法開発の研究計画の概要について報告する。

原子炉施設におけるシビアアクシデント発生時のAM<策として、水源が枯渇した後に原子炉及び使用済燃料プールに海水注入され、これがホウ酸と混合する状況において、炉心及びデブリベッドの冷却性に及ぼす影響を把握するための試験を計画している。ここでは、その構想をまとめる。

曲率を有する平面上の液膜分布を高速計測可能な液膜センサを開発し、環状流路を構成する直径12mm内管および直径18mm外管それぞれの表面に液膜センサを配置することで、液膜挙動を両面で同時に計測した。水-空気体系において、気液混合部から下流側300mmの位置に形成される液膜挙動を10&times;32計測点の液膜分布として5000断面/秒で取得し、気液混合条件および入口みかけ流速の影響を液膜厚さ，波速などによって評価した。

ワイヤメッシュセンサで大口径管内二相流を計測し、得られた信号から個々の気泡信号を識別し、それぞれの気泡の移動速度を求めた。

A wire-mesh sensor (WMS) can acquire a void fraction distribution at a high temporal and spatial resolution and also estimate the velocity of a vertical rising flow by investigating the signal time-delay of the two WMSs. The authors propose to extend this time series analysis to estimate the multi-dimensional velocity profile via cross-correlation analysis between a point of upstream WMS and multiple points downstream. Moreover, bubbles behave in various ways according to size, which is used to classify them into certain groups.

The carbon-doped copper oxide is devised as an electrocatalyst to reduce carbon dioxide under ambient pressure and temperature. The electrode was immersed into a KHCO_3 aqueous solution with CO_2 bubbling. The electric potential was maintained at -1.64 V vs SHE. The electrode prepared at 900 ℃ gives the maximum production rate of ethylene (25 %), ethanol (6.9 %) and 1-propanol (3.6 %). The production rate of methane, from which is harmful to separate ethylene, was suppressed to one fifteenth of that of ethylene. In contrast to a thermally-oxidized copper-oxide layer, the doped-carbon and a high ratio of Cu_2O to CuO in the devised electrocatalyst may result in the higher productivity and selectivity.

The best-estimate code TRACE is validated with the stability database of SIRIUS-N Facility at low pressure. The SIRIUS-N facility simulates thermal-hydraulics of the economic simplified BWR (ESBWR) Numerical results exhibits flashing-induced density wave oscillation characteristics, since the oscillation period correlates well with single-phase liquid transit time through the chimney region regardless of the system pressure, inlet subcooling and heat flux Unlike Type-I and II density wave oscillations, the inlet or exit throttling does not affect stability boundary and oscillation amplitude of flashing-induced density wave oscillations significantly. Stability maps in reference to the subcooling and heat flux obtained from the TRACE code agrees with those of the experimental data at low subcooling region, though instability observed in the lower heat flux and higher subcooling region from the stability limit of the experiment.

ジルコニウム合金の表面改質技術フレッシュグリーン(FG)は緻密で密着性が高い保護皮膜であることから、耐食性と耐水素吸収性が向上する。そのメカニズムを検討するため、三点曲げ試験および球形ナノインデンテーションを実施し、FG皮膜に対して機械的特性（亀裂発生応力、弾性率、降伏せん断応力など）を測定し、大気酸化皮膜と比較して考察した。

10×10バンドル流路内気液二相流を断面分布として高速計測可能なサブチャンネルボイドセンサを開発し、空気-水二相流動実験へ適用した。さまざまな流動条件に対して得られた実験結果に基づきボイド率分布，相速度分布，気泡コード長分布などによって二相流遷移過程を評価するとともに、既存のドリフトフラックスモデルの予測精度を検証した。

下降流では、モデル開発に必要な実験データおよび流動メカニズムに関する知見が上昇流に比べ不足している。本研究では、ワイヤメッシュセンサ多次元二相流計測手法を水空気系小口径円筒管内下降流及び上昇流計測に適用した。得られた平均ボイド率、平均気泡速度の比較から下降流と上昇流との流動特性の違いを評価することで、下降流に関する既存ドリフトフラックスモデルの適用性を検証した。

ジルコニウム合金の表面を炭化と酸化を同時に進行させる表面改質技術フレッシュグリーン(FG)は密着性が高く、皮膜が緻密で保護性が高いことから、耐食性と耐水素吸収性が向上する。さらに白金を共担持することにより、耐食性のみならず耐水素吸収性を著しく高めることが360℃高温水中腐食試験で判明した。白金・炭素共担持FG皮膜はMott-Schottky分析によりドナー密度が高く、自然浸漬電位が貴であることが水素吸収を抑制したと考えられる。

燃料被覆管のブレイクアウェイ現象は、被覆管の腐食を加速させる。ブレイクア ウェイの周期は、合金組成や水化学環境により異なり、被覆管表面に形成される。酸化膜の内部応力の違いによると推定される。酸化膜は数&mu;mの厚さのため、詳細 な応力解析には高い分解能が必要である。そこで、 放射光によるマイクロビームX線を利用した被覆管酸化膜の局所X線回折と応力解析手法を検討した。<br> X線回折に寄与する体積を最小化するため、ビーム幅を0.2&mu;mとして、高温水中腐食により形成した酸化膜を深さ方向に透過法を用いてX線回折測定した。単斜晶酸化ジルコニウムの格子面間隔は、基材との界面より表面側へ1&mu;mの付近から顕著に狭く、圧縮応力は膜内部の方が界面付近よりも大きいことを示唆している。これは、正方晶酸化ジルコニウムのマルテンサイト変態による体積膨張が圧縮応力と して作用したためと考えられる。

10&times;10バンドル流路内気液二相流を断面分布として高速計測可能なサブチャンネルボイドセンサを開発し、入口液相みかけ流速が0であるプール体系の空気-水二相流動実験へ適用した。実験結果からボイド率分布，相速度分布，気泡コード長分布の軸方向発達過程を評価することにより、既存モデルの予測精度を検証した。

The temperature dependence on radiation induced surface activation (RISA) corrosion control effect for stainless steel was examined in artificial and natural seawater in this paper. The results obtained are as follows: 1) The RISA effect was evident up to 30 °C, but was hardly recognizable over 45 °C. 2) The RISA effect was observed regardless of the seasons, and the anodic current decreased with increase in the temperature. 3) It was confirmed that corrosion control mechanism proposed in the previous paper was appropriate. The increase of dissolved oxygen was measured in natural seawater by the appearance of RISA effect.

The temperature dependence on radiation induced surface activation (RISA) corrosion control effect for stainless steel was examined in artificial and natural seawater in this paper. The results obtained are as follows: 1) The RISA effect was evident up to 30 °C, but was hardly recognizable over 45 °C. 2) The RISA effect was observed regardless of the seasons, and the anodic current decreased with increase in the temperature. 3) It was confirmed that corrosion control mechanism proposed in the previous paper was appropriate. The increase of dissolved oxygen was measured in natural seawater by the appearance of RISA effect.

A wire-mesh sensor (WMS) can acquire a void fraction distribution at a high temporal and spatial resolution and also estimate the velocity of a vertical rising flow by investigating the signal time-delay of the two WMSs. The authors propose to extend this time series analysis to estimate the multi-dimensional velocity profile via cross-correlation analysis between a point of upstream WMS and multiple points downstream. Moreover, bubbles behave in various ways according to size, which is used to classify them into certain groups.

ワイヤを正方格子状に配したワイヤメッシュセンサ（Wire-Mesh Sensor: WMS）は、ワイヤが交差する近接点の導電率から気泡の有無の二次元分布を高速・高精度で計測である。さらに、WMSは近接配置した2組の計測信号の時間差により気泡速度も計測可能である。本研究では、大気圧水‐空気条件において大口径円管内二相流動試験を行い、WMSを二相流動計測に適用し、ボイド率および気泡径の二次元分布計測、気泡速度の三次元計測を行った。気泡速度の三次元計測は上流側WMSの１計測点に対し下流側WMSのある計測点領域（Nx×Ny）間で相互相関分析することにより行った。さらに、気泡信号を気泡径ごとに弁別、抽出し、気泡の大きさごとの三次元速度計測も行うことで、大口径円筒管における二相流動の発達過程を気泡サイズ毎に評価した。

ワイヤメッシュセンサ(WMS)はコンダクタンス法に基づき断面ボイド率分布を計測でき、近接した2層目のWMSとのボイド信号の相互相関により気泡速度を推定出来る。本研究では多次元速度分布を推定するため、大口径円管内を上昇する二相流動を対象に、上流側WMSの１計測点と、下流側の21×21計測点との間で相互相関分析を行い、相互相関係数が最大になる組み合わせで速度ベクトルが定めた。さらに、WMS信号のWavelet解析を行い、気泡信号を周波数（気泡の大きさ）ごとに抽出し、気泡の大きさ毎の多次元速度を求めた。本手法を旋回流に対して適用した結果、気泡の旋回成分が計測され、大気泡は円筒管中心へ移動することを定量的に評価した。

管群流路内の気液二相流動を多次元的に高い時間分解能で計測可能なセンサを開発した。開発したセンサは、10×10本の管群流路において各管のギャップに11×11本のワイヤ電極ならびに100本のロッド電極により構成され、各電極間のコンダクタンス計測結果に基づきボイド率を推定する。隣接する管に囲まれた流路であるサブチャンネルの中央に対応するワイヤ同士の近接点121箇所(11×11)、および管ギャップに対応する管とワイヤの近接点400箇所(100×4)の二相流計測が可能である。水空気二相流動試験により、管群流路内の軸方向ならびに径方向ボイド混合特性をボイド率分布、相速度分布などによって定量的に評価した。

The authors developed a method of measurement to determine the multi-dimensionality of two phase flow. A wire-mesh sensor (WMS) can acquire a void fraction distribution at a high temporal and spatial resolution and also estimate the velocity of a vertical rising flow by investigating the signal time-delay of the upstream WMS relative to downstream. Previously, one-dimensional velocity was estimated by using the same point of each WMS at a temporal resolution of 1.0-5.0s. The authors propose to extend this time series analysis to estimate the multi-dimensional velocity profile via cross-correlation analysis between a point of upstream WMS and multiple points downstream. Bubbles behave in various ways according to size, which is used to classify them into certain groups via wavelet analysis before cross-correlation analysis. This method was verified by air-water straight and swirl flows within a large-diameter vertical pipe. A high-speed camera is used to set the parameter of cross-correlation analysis. The results revealed that for the rising straight and swirl flows, large scale bubbles tend to move to the center, while the small bubble is pushed to the outside or sucked into the space where the large bubbles existed. Moreover, it is found that this method can estimate the rotational component of velocity of the swirl flow as well as measuring the multi-dimensional velocity vector at high temporal resolutions of 0.2s.

自然循環ループにおける密度波振動には、水頭損失が支配的なI型と、摩擦損失が支配的なII型に分類される。本研究では、自然循環沸騰水型原子炉の一つであるESBWRを摸擬したSIRIUS-N設備で観測したI型密度波振動を対象に、システム動特性解析コードTRACE ver5を用いた数値解析を行った。その結果、解析では実験で得られた静特性および安定境界を妥当な精度で予測した。また不安定現象の特徴である振動周期と通過時間の相関を精度良く予測できることから、TRACEコードが自然循環ループの安定性評価に使用することの有用性を確認した。

大口径円筒管内の二相流動は多次元性を持つため、多次元二相流計測手法は気泡流やチャーン流のモデル化において重要な計測手法である。ワイヤメッシュセンサ（WMS）はボイド率分布を高速かつ高空間分解能で計測可能である。これまで二段のWMSにより気泡のWMS通過時間差から主流方向一次元相速度計測も行われている。本研究では、大口径円筒管（内径：224mm）内二相流動をWMSにより計測した。上流側WMSのある点の信号と、下流側WMS複数点の信号との間で相互相関分析を行い、最も高い相関係数を示す方向と時間差から多次元速度ベクトルを求めた。また、小気泡は液相流動による影響が大きく、大気泡は多次元せん断流となる。多次元相速度ベクトルは気泡径により大きく異なるため、Wavelet解析を用いて気泡径毎に信号を抽出し、気泡径毎に相速度ベクトルを評価した。

ジルコニウム合金の表面改質技術フレッシュグリーン(FG)は緻密で密着性が高い保護皮膜であることから、耐食性と耐水素吸収性が向上する。そのメカニズムを検討するため、FG皮膜の電気化学特性を試験した。基材にジルカロイ2を用い、対照試験として、FG皮膜の他に大気中で酸化した(OD)皮膜と360℃高温水中で酸化させた(AC)皮膜を用いた。EISの結果、ACとOD皮膜は利得も位相遅れもほぼ同じであるが、FG皮膜はコンダクタンスが高い二層構造から構成されていることが判明した。Mott-Schottky分析によりFG皮膜はACとOD皮膜よりドナー密度が高い。さらに自然浸漬電位が高い。これらは酸化皮膜内にカーボンがドープされていることが原因と推察される。これらの電気化学特性がFG処理被覆管材の耐食性を向上させていると考えられる。

10×10バンドル燃料集合体を模擬した流路内の二相流動の三次元分布を高い時間分解能で計測可能なセンサを開発した。開発したセンサは、模擬燃料棒ギャップに配置した11×11本のワイヤ電極ならびに模擬燃料棒である100本のロッド電極により構成され、各電極間のコンダクタンス計測結果に基づきボイド率を推定する。サブチャンネル中央に対応するワイヤ同士の近接点121箇所(11×11)、および燃料棒ギャップに対応する模擬燃料棒とワイヤの近接点400箇所(100×4)の二相流計測（ボイド率、気泡コード長、相速度）が可能である。水空気二相流動試験により、燃料集合体内の鉛直上昇二相流動を定量的に計測できることを検証した。

ジルコニウム合金の表面改質技術フレッシュグリーン(FG)は緻密で密着性が高い保護皮膜であることから、耐食性と耐水素吸収性が向上する。そのメカニズムを検討するため、FG皮膜のラマン分光分析を行った。基材にジルカロイ2を用い、対照試験として、FG皮膜の他に大気中で酸化した(OD)皮膜と360℃高温水中で酸化させた(AC)皮膜を用いた。ラマン分光分析の結果、フレッシュグリーン皮膜では、ドープされる炭素の一部はsp2軌道を有する非晶質炭素の形態であると推定される。フレッシュグリーン皮膜表面から約0.5umの深さに於いて、Gバンドの強度が平均値の3倍以上になるピークが観測されている。フレッシュグリーン皮膜では、表面では単斜晶の割合が大きく、0.6umより深い位置から、基材との界面近傍までにおいて正方晶の割合が相対的に高くなる領域が存在している。この境界はGバンドのピークと符合している。これらの電気化学特性がFG処理被覆管材の耐食性を向上させていると考えられる。

バンドル流路内気液二相流を高速で三次元計測可能なボイドセンサを開発した。開発したセンサは、10×10バンドル流路において隣接するロッドに囲まれたチャンネルの中央サブチャンネル121箇所(11×11)、およびロッド表面サブチャンネル400箇所(10×10×4)を複数の流路断面に対して同時計測可能である。空気-水二相流動実験により取得したボイド率分布，相速度分布，気泡コード長分布に基づき既存モデルと比較することで、バンドル内気液二相流の発達特性を評価した。

大口径管内やバンドル流路などの複雑流路を対象として、液膜厚さ分布測定法や三次元界面速度測定法を開発した。<br>(1)液膜厚さ分布を1ms毎に計測するセンサーとして、励起電極と計測電極の対を縦横1.8mm間隔で3層プリント基板に実装した。直径12mmの円筒状に曲げ加工し、円筒表面の液膜流動を可視化した。<br>(2)広く使用されているワイヤメッシュセンサの計測値から気液構造の大きさ毎に三次元速度分布を推定する方法を考案した。旋回流に適用した結果、遠心力により液相が外側に移行し、気相が内側に向かう運動を可視化した。<br>(3)10本×10本の円管バンドルの間隙に細線を縦に11本、横に11本挿入し、サブチャンネル中央121点を可視化した。さらに円管を励起電極に、細線を計測電極にすることで、円管ギャップの400点を可視化した。主流方向に8断面を521点ずつ1ms毎に計測するサブチャンネルボイドセンサーを開発した。

ジルカロイ-2の表面を薄いカーボンドープ酸化ジルコニウム層に表面改質するフレッシュグリーン処理を施した。ラマン分光分析を用いて分析を行った結果、フレッシュグリーン皮膜では、GバンドとDバンドに特徴的な共鳴ピークが現れ、カーボンはZrC結合以外にもsp2軌道を有するグラファイト構造でドープされていることが判明した。

ジルカロイ-2の表面を薄いカーボンドープ酸化ジルコニウム層に表面改質するフレッシュグリーン処理を施した。SPring-8放射光施設においてマイクロX線を用いてフレッシュグリーン皮膜の結晶構造を解析した結果、皮膜の酸化ジルコニウムの結晶構造は表面から金属との界面にかけて単斜晶であり、界面近傍では正方晶であることが明らかとなった。

著者らは液膜流動を高い時間・空間分解能で計測する高密度多点電極法を開発した。センサーはn行m列の電極対で構成され、それぞれn行の電圧印加とm列の電流走査でn+m本の電極リードにまとめることで高密度化を達成している。液膜厚さの計測原理は電極間のコンダクタンス計測であるため、電圧印加時に形成される電気二重層の特性を知ることが重要である。本研究では、電気化学インピーダンス法により等価回路を求め、電気二重層の静電容量や溶液抵抗などを推定した。センサーの過渡応答により計測速度の制限値などを見出した。

高温ステンレス球をナノ流体および塩水に浸漬するクエンチ実験を行った。塩水では水溶液濃度の増大に伴いクエンチ温度が上昇したが、ナノ流体ではその傾向は顕著に見られなかった。この違いを考察するために、常温常圧で気液界面からの蒸発速度を観測する実験を行った。その結果、塩水では濃度の増大に伴い蒸発速度が減少したが、ナノ流体では水とほぼ変わらない値であった。よって、ナノ流体と塩水とのクエンチ温度の変化は蒸発速度と相関があることが明らかになった。

ワイヤメッシュセンサでは断面のボイド率分布や流速分布を連続的に計測できるが、侵襲法であるため二相流構造に影響を及ぼす。本研究では内径50mmの円管を用い、気泡上昇の可視画像と対比することにより、様々な気泡径および気泡形状に対してワイヤメッシュセンサが界面変形に与える影響を明らかにした。また、気泡上昇速度の計測精度を気泡径および径方向測定位置の関数として表し、二相流計測手法としてのワイヤメッシュセンサの適用性を把握した。

液膜流動を高い時間・空間分解能で計測する手法を開発し、液膜厚さの過渡変化が計測可能であることを示した。本手法は、曲率を有する構造体として燃料被覆管を模擬した体系への適用も可能である。

ワイヤメッシュセンサ（WMS）は断面ボイド率分布や界面移動速度分布を高速に計測できる。本研究では、内径224mmの配管内鉛直上昇流におけるWMSの計測精度を高速度カメラ等との比較により検証した。

燃料棒を模擬した円柱形状表面の液膜流動を2次元分布として高い時間分解能で計測可能な高密度多点電極センサを開発した。開発したセンサを用いて水空気による液膜流動試験を実施することにより、模擬燃料棒表面における垂直上昇液膜流動を計測できることを示した。

内径224mmの円筒管内垂直上昇気液二相流を対象に数値流体解析(CFD)を行った。CFDには汎用流体解析コードを用いた。実験で得られたボイド率や気泡上昇速度の径方向分布を解析結果と比較して精度を検証した。また、二相流相関式のパラメータの感度を検討した。

炭素ドープと酸化処理を同時に施す新たな表面改質処理(フレッシュグリーン処理)を施したチタンのガス放出特性と真空中の摩擦特性について調べた結果を報告する．

直径30mmの高温固体ステンレス球を、ナノ粒子を懸濁させた水、いわゆるナノ流体中に浸漬させ、膜沸騰蒸気膜の崩壊挙動を観測した。固体球表面に埋め込んだ熱電対による固体球表面温度を計測および膜沸騰蒸気膜挙動の可視観測を実施した。各沸騰様相における固体球と水の接触を検出するために、ナノ流体と固体球に電極を挿入し、固体球とナノ流体の間を流れる電流を計測した。ナノ粒子の種類および濃度を変化させることによってナノ粒子懸濁の影響を評価した。

沸騰水型原子炉(BWR)の炉内流動を模擬した試験設備SIRIUS-Fを用いて、出力および流量が減少する過渡事象を模擬する実験を行った。実験の結果、出力の異なる並行2チャンネルの流量配分は、初期は出力が高いチャンネルの方が質量流束が低い。しかしながらボイド率がある程度低減すると、質量流束が逆転する。BWR過渡事象解析コードTRAC-BF1では、この質量流束逆転現象を精度良く再現した。このような出力・流量急減事象においても、同コードはBWR炉内の沸騰二相流を精度良く予測できることを確認した。

Flow Accelerated Corrosion (FAC) requires considerable attention in plant piping management, for its potential of catastrophic pipe rupture of main piping systems. In view of fluid dynamics, the most essential factor to be considered is mass transfer at the inner surface of the pipe. Mass transfer coefficients are determined by fluid properties and piping geometry, however, no universal correlation exists, which is adaptable to various types of piping elements with strong turbulence. In this study, the modeling of mass transfer coefficient was progressed based on Chilton-Colburn analogy and utilizing "effective friction velocity" from the hydraulics in the viscous sub-layer along the wall. FAC experiments with PWR condensate water condition and CFD for the flow were conducted with a contracted rectangular duct. By considering the turbulent velocity of the viscous layer into the mass transfer coefficient, the correlation with the FAC thinning rate improved, effectively.

既存の高温固体球浸漬実験から得られた知見を基に、膜沸騰蒸気膜の線形安定性解析の定数を定め、熱伝達係数や熱物性値が蒸気膜崩壊条件に及ぼす影響を考察した。

ジルカロイ-2の表面を薄いカーボンドープ酸化ジルコニウム層に表面改質するフレッシュグリーン処理を施した。360℃飽和水中に153日間浸漬した結果、腐食増量および水素吸収量が低減する結果を得た。

A high-temperature stainless-steel sphere was immersed into various salt solutions to test film boiling behavior at vapor film collapse. The film boiling behavior around the sphere was observed with a digital-video camera. Because salt additives enhanced condensation heat transfer, the observed vapor film was thinner. Surface temperature of the sphere was measured. Salt additives increased the quenching (vapor film collapse) temperature, because frequency of direct contact between sphere surface and coolant increased. Quenching temperature rises with increased salt concentration. The quenching temperature is well correlated with ion molar concentration, which is a number density of ions, regardless of the type of hydrated salts.

When a metal oxide is irradiated by gamma rays, the irradiated surface becomes hydrophilic. This surface phenomenon is called as radiation-induced surface activation (RISA) hydrophilicity. In order to investigate gamma ray-induced and photoinduced hydrophilicity, the contact angles of water droplets on a titanium dioxide surface were measured in terms of irradiation intensity and time for gamma rays of cobalt-60 and for ultraviolet rays. Reciprocals of the contact angles increased in proportion to the irradiation time before the contact angles reached its super-hydrophilic state. The irradiation time dependency is equal to each other qualitatively. In addition, an effect of ambient gas was investigated. In pure argon gas, the contact angle remains the same against the irradiation time. This clearly indicates that certain humidity is required in ambient gas to take place of RISA hydrophilicity. A single crystal titanium dioxide (100) surface was analyzed by X-ray photoelectron spectrometry (XPS). After irradiation with gamma rays, a peak was found in the O1s spectrum, which indicates the adsorption of dissociative water to a surface 5-fold coordinate titanium site, and the formation of a surface hydroxyl group. We conclude that the RISA hydrophilicity is caused by chemisorption of the hydroxyl group on the surface.

Experiments were conducted in which a molten copper droplet released into water pool. Spontaneous vapor explosion did not occur when water temperature was over 50℃. Spontaneous vapor explosion did, however, occur when water temperature was lower than 50℃. A high-speed video frames explored the stages of vapor explosions: (1) a vapor film was formed and separates the copper droplet and surrounding water, (2) a filament of molten copper grew from the surface and deformed the vapor film, (3) the vapor film collapsed along the filament surface, and finally (4) triggering of vapor explosions occurred from the filament to the whole molten copper droplet. When the filament growth was observed, it triggered the vapor explosion in almost all .cases. When.. not, . vapor explosion was not observed and the vapor film was, therefore, stably formed around the molten copper droplet. We concluded that the filament form the molten copper triggered vapor explosion in a highly subcooled water.

The fine-scale mixing morphology of vapor explosion and droplet entrapping was visualized by high-speed digital video cameras at an interface between a water droplet and a molten tin pool. The cluster of bubble generated by spontaneous bubble-nucleation covered the whole contact area at 0.1ms after the impingement. Prominent fine mixing between two liquids were found to start at 0.6ms that resulting in vapor explosion. Droplet entrapping phenomenon frequently occurred on an oxide layer, since coherent mixing was prevented due to unevenly formed oxide layer.

A high-temperature stainless-steel sphere was immersed into an A12O3 nanofluid to investigate film boiling heat transfer and collapse of vapor film. Surface temperature is referred to the measured value of thermocouples embedded into and welded onto bottom surface of the sphere. The Al_2O_3 nanofluid concentration is varied from 0 to 15 wt%. Comparing between Al_2O_3 nanofluid and salt solution, The film boiling heat transfer coefficient and quenching temperature for Al_2O_3 nanofluid remain the same within the experimental Al_2O_3 nanofluid concentration. In contrast to the nanofluid, those values for salt solution significantly increased with increasing concentration of the salt solution.

In order to store nuclear spent fuels for a long term, we propose the concept of stainless steel canister with titanium cladding. A parametric study on MIG brazing for titanium and stainless steel revealed that Cu-1Mn-3Si alloy (MG960) is promising MIG brazing material. The optimized brazing conditions showed adequate structural strength such as JIS G 0601 shear strength, tensile shear stress and peel strength.

SUS304鋼上にAlO3溶射被膜やプラズマ溶射被膜を成膜した試験片にRI放射線源によりγ線を照射しながら、この試験片の腐食溶液中での母材の溶出量および電気化学的変化を計測した。これらの条件で、放射線誘起表面活性によるSUS304鋼の隙間腐食抑制効果を確認した。

塩の析出および粒子懸濁が膜沸騰蒸気膜の崩壊挙動に及ぼす影響を明らかにするため、水溶液中に直径30mmの高温固体ステンレス球を浸漬させる蒸気膜崩壊実験を実施した。固体球表面温度の計測ならびに膜沸騰蒸気膜挙動の可視観測を実施した。各沸騰様相における固液接触による固体球と水の接触を検出するために、水溶液と固体球に電極を挿入し、固体球と水溶液の間を流れる電流を計測した。塩の析出の影響については、あらかじめ塩を析出させた固体球を用いた蒸気膜崩壊実験により評価し、粒子懸濁の影響については、懸濁水溶液を用いた蒸気膜崩壊実験により評価した。

800℃に加熱した高温固体ステンレス球を80℃の水和塩水溶液に浸漬する実験を実施し、固体球周囲に形成される膜沸騰蒸気膜挙動および固体球表面への水和塩析出挙動を可視観測した。塩が析出しやすい水溶液および塩が析出しにくい水溶液に同様の実験を実施し、膜沸騰蒸気膜の崩壊挙動と水和塩析出挙動の相関を明らかにした。

SIRIUS-F設備を用いて、BWRの流量急減事象を模擬した実験を行った。流量および出力が急減する場合においても、出力の異なる2並行流路（模擬炉心チャンネル）の出入口間の差圧は同じ値を示した。実験で得られた流量や差圧等の沸騰二相流挙動は、過渡時においても数値解析と良い一致を見せた。

多孔質皮膜としてジルコニア熔射皮膜を鉄母材に成膜し、0.1%NaCl水溶液中でコバルト60からのγ線を照射して防食性能を調べた。溶出した鉄の量からγ線照射による防食効果が認められた。

配管減肉に対する流体現象の寄与について説明した．流れ加速型腐食（FAC）とエロージョンでは，壁面への流体作用力が全く異なることを示した．FACの主要因子である乱流物質移動は熱伝達とアナロジがあり，壁面近傍の粘性底層内の分子輸送が支配的とするモデルが実用的である．さらに物質移動は数値流体解析コードで求められる．最後にこれらの手法によりオリフィス，エルボにおける物質移動現象を予測した結果を説明した．

鉄母材上にジルコニア溶射皮膜を成膜した試験片にγ線を照射しながら、この試験片の腐食溶液中での母材の溶出量および電気化学的変化を計測した。このとき、放射線誘起表面活性(RISA)によると考えられる効果が見いだされた。この結果からRISA効果の発現メカニズムについて新たな知見を見いだした。

水溶液中に直径30mmの高温固体ステンレス球を浸漬させる蒸気膜崩壊実験を実施した。固体球表面温度の計測ならびに膜沸騰蒸気膜挙動の可視観測を実施した結果、蒸気膜崩壊温度に及ぼす水溶液種類および濃度の影響を、水溶液中のイオン数密度を表すイオンモル濃度によって統一的に整理できることを示した。

ジルカロイ-2の表面を薄いカーボンドープ酸化ジルコニウム層で覆う表面改質処理（フレッシュグリーン処理）を施した。高温蒸気中における水素吸収の改善効果を調べた結果、水素吸収量が大幅に低下するとの結果を得た。

チタンの表面を薄いカーボンドープ酸化チタン層で覆う表面改質処理（フレッシュグリーン処理）を施した。耐食性の改善効果を調べた結果、耐食性が問題とされていた1%沸騰塩酸や30%沸騰ギ酸などに対して、全面腐食耐性が大きく改善された。

4A族であるチタン、ジルコニウム、およびハフニウムの表面を改質処理する技術（フレッシュグリーンと命名）を開発した。この処理によって、表面は薄い酸化物皮膜で覆われるが、その皮膜には炭素が含まれる。このカーボンドープ酸化物皮膜は緻密で密着力が強く、高い硬度、耐食性や耐摩耗性が発現する。

This study examines a corrosion control technique for corrosion-resistant materials or of stainless steel. This employs an effect of Radiation Induced Surface Activation (RISA) . The experimental results revealed: (1) The mechanism behind the corrosion control proposed by the previous report was confirmed to be appropriate. This via tests that measured the amount of dissolved oxygen and iron ions, in the solution. (2) The corrosion control technique was confirmed to be useful for stainless steel with any kind of metal oxide film coating on the surface. (3) It was also shown to be useful even in actual seawater, due to biological effects, which is a far more severe environment for corrosion control than simple salt water. The corrosion control technique for corrosion-resistant material using RISA in seawater has therefore been shown to offer a significant potential for practical applications in naval architecture and marine structures.

In order to investigate the stability of a nuclear reactor core with an oxide mixture of uranium and plutonium (MOX) fuel installed, channel stability and regional stability tests were conducted with the SIRIUS-F facility. The SIRIUS-F facility was designed and constructed to provide a highly accurate simulation of thermal-hydraulic (channel) instabilities and coupled thermalhydraulics-neutronics instabilities of the Advanced Boiling Water Reactors (ABWRs). A real-time simulation was performed by modal point kinetics of reactor neutronics and fuel-rod thermal conduction on the basis of a measured void fraction in a reactor core section of the facility.<br>A time series analysis was performed to calculate decay ratio and resonance frequency from a dominant pole of a transfer function by applying auto regressive (AR) methods to the time-series of the core inlet flow rate. Experiments were conducted with the SIRIUS-F facility, which simulates ABWR with MOX fuel installed. The variations in the decay ratio and resonance frequency among the five common AR methods are within 0.03 and 0.01 Hz, respectively. In this system, the appropriate decay ratio and resonance frequency can be estimated on the basis of the Yule-Walker method with the model order of 30.

Channel stability, core-wide stability, and regional stability experiments were conducted for a wide range of operating conditions of a BWR-5 with UOX (uranium dioxide) type 9&times;9A fuels installed, including maximum power points along the minimum pump speed line and the natural circulation line. The SIRIUS-F facility, used in this study, was designed and constructed for highly accurate simulation of void-reactivity feedback of BWRs. The decay ratios and the resonance frequencies are in good agreement with those from the design analysis code, ODYSY. The SIRIUS-F experimental results demonstrated stability characteristics such as a stabilizing effect of the power, and revealed a sufficiently large stability margin even under hypothetical conditions of power enlargement.

Flow Accelerated Corrosion (FAC) requires considerable attention in plant piping management, for its potential of catastrophic pipe rupture of main piping systems. In view of fluid dynamics, the most essential factor to be considered is mass transfer at the inner surface of the pipe. Mass transfer coefficients are determined by fluid properties and piping geometry, however, no universal correlation exists, which is adaptable to various types of piping elements with strong turbulence. In this study, the modeling of mass transfer coefficient was progressed based on Chilton-Colburn analogy and utilizing "Effective Friction velocity" from the hydraulics in the viscous sub-layer along the wall. FAC experiments with PWR condensate water condition and CFD for the flow were conducted with a contracted rectangular duct. By considering the turbulent velocity of the viscous layer into the mass transfer coefficient, the correlation with the FAC thinning rate improved, effectively.

In order to investigate the stability of a nuclear reactor core with an oxide mixture of uranium and plutonium (MOX) fuel installed, channel and regional stability tests were conducted with the SIRIUS-F facility. The SIRIUSF facility was designed and constructed to provide a highly accurate simulation of thermal-hydraulic (channel) instabilities and coupled thermalhydraulics-neutronics instabilities of the Advanced Boiling Water Reactors (ABWRs). A realtime simulation was performed by modal point kinetics of reactor neutronics and fuel-rod thermal conduction on the basis of a measured void fraction in a reactor core section of the facility. A time series analysis was performed to calculate decay ratio and resonance frequency from a dominant pole of a transfer function by applying AR methods to the time-series of the core inlet flow rate. Experiments were conducted with the SIRIUS-F facility, which simulates ABWR with MOX fuel installed. The variations in the decay ratio and resonance frequency among the five common AR methods are within 0.03 and 0.01 Hz, respectively. In this system, the appropriate decay ratio and resonance frequency can be estimated on the basis of the Yule-Walker method with the model order of 30.

1500℃の溶融錫を内径2mmのノズルから常温の冷却材プール中に連続滴下させる実験を、冷却材として水ならびに5、10、20wt%塩化カルシウム水溶液を用いて実施した。その結果、水の場合には安定した膜沸騰形成のためにほとんど蒸気爆発することなく固化に至ったが、塩化カルシウム水溶液の場合には水溶液濃度の上昇とともに蒸気爆発頻度が増大した。得られた固化物の粒径分布を比較すると、水の場合には9割以上が直径1mm以上であったのに対して、20wt%塩化カルシウム水溶液の場合には8割以上が直径1mm以下となり、上限付対数正規分布とよい一致がみられた。

ステンレスの表面に酸化チタンの薄い皮膜を成膜し、電気化学的に酸化チタンの自然浸漬電位を測定した。この皮膜にγ線もしくは紫外線を照射した場合の電位応答を調べた。酸化チタンの皮膜はスパッタ成膜であり、厚さは 100nmから1000nmである。この厚さの範囲では、短波長紫外線(UV-C)照射による電位変化および応答の速さは皮膜厚さに依らない傾向を示した。これは半導体としての酸化チタン皮膜の応答として説明できるものである。一方、γ線を照射した場合には、自然浸漬電位の値が皮膜厚さに対して相関しておらず、振る舞いは単純でないが、皮膜がステンレス上に成膜された複合材料であることを考慮に入れて定性的な説明を与える。

A corrosion mitigation technique based on radiation induced surface activation (RISA) from the gamma ray irradiation on a metal surface is reported in this paper. This study aimed to develop a RISA method to prevent crevice corrosion in SUS304 stainless steel using low-intensity radioactive material. Experiment showed that an electrode potential of -100 mV vs. Ag/AgCl was produced and maintained on TiO₂-coated SUS304 stainless steel specimens immersed in artificial seawater and in close contact with a small, sealed⁶⁰Co source or activated by spontaneous neutron irradiation, with no corrosion observed for more than 7 days. On the contrary, the potential of the specimen without a radiation source decreased less than-280 mV vs. Ag/AgCl and crevice corrosion occurred beneath the O-ring within few days. The RISA effect of low-intensity radioactive material has the potential to prevent crevice corrosion of SUS304 stainless steel in actual seawater.

放射線誘起表面活性（Radiation Induced Surface Activation, RISA）は、今世紀初頭に我が国で初めて確認された金属酸化皮膜への放射線照射によって生じる表面活性効果であり、その効果により放射線環境下の伝熱・防食特性向上をはかることができる。本稿では、昨年度末まで4年計画で行われた経産省革新的実用原子力技術開発費補助事業「放射線誘起表面活性効果による高性能原子炉に関する技術開発」のRISAによる伝熱特性向上及びメカニズム解明などの研究結果を概説する。

使用済燃料をコンクリートキャスク等で貯蔵する場合の長期耐久性に優れたキャニスターを提案する。通常のステンレス製キャニスターをチタン板で被覆する構造である。これによりステンレス製キャニスターの腐食や応力腐食割れの懸念が無くなる。従来、チタンとステンレス鋼の溶接においては、脆い金属間化合物が析出し、強度が不足することが知られてきたが、銅合金を用いることによってチタンとステンレス鋼との溶接を行い、接合部は十分な強度があることが判明した。コスト評価を実施した結果、本概念ではチタン使用量が少なく、低コストで製造できる見通しを得た。なお、本キャニスターの概念は、放射性廃棄物を処分する場合などで用いる容器の長期耐久性確保にも有効である。

高温固体ステンレス球を冷却材プールに浸漬させ、高温固体球周囲に形成される膜沸騰蒸気膜を可視観測した。膜沸騰時の可視観測画像から蒸気膜厚さおよび凝縮熱伝達率を算出した結果、水和塩の添加によって蒸気膜厚さが減少し、凝縮熱伝達が向上することが判明した。

We proposed ultra rapid solidification and atomization technique, CANOPUS (Cooling and Atomizing based on NOble Process Utilizing Steam explosion), using small-scale vapor explosions to make an amorphous metal. The CANOPUS method is suitable for rapid cooling and atomization process, which utilizing sustainable small-scale vapor explosions. In order to apply the CANOPUS method to a high melting point metal, it is necessary to make a small-scale vapor explosion occur at a high temperature of the molten metal. Small-scale experiment is conducted to develop the vapor explosion promotor in which spontaneous vapor explosion can occur at a high temperature of a molten metal. Spontaneous vapor explosion do not occur when water at 80&deg;C is used as a coolant. However, spontaneous vapor explosion occurs when water at 80&deg;C with salt additives is used as a coolant. Specifically, lithium chloride solution generates spontaneous vapor explosions at the highest temperature of the molten tin in the experiment. In order to clarify the triggering mechanism of the spontaneous vapor explosion when the promotor is used as a coolant, a high-temperature solid stainless steel sphere is immersed into a coolant. The interfacial temperature of the stainless steel sphere is measured, and the behavior of a vapor film around the stainless steel sphere is observed with a digital video camera. As a result, salt additives resulted in an increase of quench temperature in all salt solutions. The quenching curves of each coolant indicate that the salt additives improve the film boiling heat transfer. The improvement of the film boiling heat transfer causes an unstable formation of the vapor film and a rise of the quench temperature. It is clarified that the salt additives to water promotes a vapor film collapse. Comparing two experiments, the quench temperature of each solution is in close agreement with the upper limit of the molten tin temperature that causes spontaneous vapor explosion. This result suggests that the vapor film collapse triggers spontaneous vapor explosion.

酸化金属表面に放射線（γ線）を照射することにより皮膜表面が親水化する現象が見出されている。本報告では、照射による親水化面の安定性に関する結果を中心に報告する。

1400度の銅液滴を水中に滴下する実験を行い、蒸気爆発が発生する場合には、溶融銅のフィラメントが観測され、その後のトリガリングの基点になることを明らかにした。

ステンレス鋼表面に高温大気中で酸化皮膜を設け、γ線照射を行い、自然浸漬電位を測定した。その結果、電位は100mV以上卑化し、大気中酸化皮膜でも放射性誘起表面活性による腐食緩和効果が発現することが判明した。

錫を用いた小規模自発的蒸気爆発実験ならびに高温固体球を用いた蒸気膜生成・崩壊実験を行った。その結果、塩の添加によってクエンチ温度が上昇し、蒸気膜が崩壊しやすくなることで、より高温で蒸気爆発が発生するようになることを示した。この蒸気爆発促進効果は、塩添加によって膜沸騰熱伝達が向上し、生成される蒸気膜が薄くなったことに起因することを明らかにした。

BWR-5に9×9A型のMOX燃料を全体の1/3装荷した炉心を想定し、領域安定性試験設備SIRIUS-Fを用いて最低ポンプ速度最大出力点など広い範囲で試験を行い、運転範囲において、十分な安定性余裕があることを確認した。また許認可解析コードODYSYの減幅比(DR)および共振周波数は試験結果と精度良く一致した。

金属酸化物にガンマ線照射をすると放射線誘起表面活性（ＲＩＳＡ）により超親水性が発現すると考えられる。ガンマ線照射した酸化チタン単結晶(100)面を表面分析し、超親水性と表面状態の関係を示す。親水性を発現した酸化チタン単結晶表面ではブリッジ酸素が減少していることがＸＰＳからわかった。酸化チタンの光触媒としての詳細な研究から、欠損したブリッジ酸素には水が化学吸着して表面水酸基を形成すると考えられており、表面水酸基が超親水化に重要である。一方、昇温脱離分析を行ったところ、表面からの水脱離温度が高い値を示した。このことから、結合エネルギーの高い表面吸着水の存在が超親水性発現に重要であることがわかった。

Experiments were conducted in which a molten copper droplet released into water pool. Spontaneous vapor explosion did not occur when water temperature was 50℃. Spontaneous vapor explosion did, however, occur at a rate of 70%, when water temperature was 20℃. A high-speed video frames explored the stages of vapor explosions: (1) a vapor film was formed and separates the copper droplet and surrounding water, (2) a filament of molten copper grew from the surface and deformed the vapor film, (3) the vapor film collapsed along the filament surface, and finally (4) triggering of vapor explosions occurred from the filament to the whole molten copper droplet. When the filament growth was observed, it triggered the vapor explosion in almost all cases. When not, vapor explosion was not observed and the vapor film was, therefore, stably formed around the molten copper droplet. We concluded that the filament form the molten copper triggered vapor explosion in a highly subcooled water.

When a semiconductor film is irradiated by gamma rays, excited electrons are transferred to abase metal in contact with the film, resulting in a drop of corrosion potential. The authors propose a corrosion mitigation method based on radiation induced surface activation (RISA) phenomena by supplying gamma rays from outside the material, or based on a self-excited methodology activating the film and/or the base metal. The corrosion potential of ZrO₂ coated SUS304L was shifted down to the range between -90 mV and -300 mV vs. SSE by gamma-ray irradiation. The corrosion potential was further shifted down to -600 mV when a CoCr intermediate layer was inserted between the ZrO₂ spray coating film and the SUS304L base metal. Iron specimens with a spray coating film of TiO₂, ZrO₂, and Al₂O₃ were immersed in a 3 wt% sodium chloride aqueous solution. Pitting and general corrosion were observed on both the specimens kept in a darkroom and illuminated with ultraviolet rays. Pitting and general corrosion were, however, suppressed on all three specimens irradiated with gamma rays.

The present paper concerns the evaluations of the long period super-hydrophilicity of titanium oxide films fabricated by plasma thermal spraying on the aluminum substrate. The fabricated titanium oxide and tungsten doped titanium oxide kept super-hydrophilic surface over 315 days without UV ray illumination. Especially, tungsten doped titanium oxide kept super-hydrophilic surface including the area without dropping of water for 315 days.

A time series analysis was performed to calculate decay ratio and resonance frequency from a dominant pole of a transfer function by applying AR methods to time-series of the core inlet flow rate. Experiments were conducted with the SIRIUS-F facility, which simulates ABWR with MOX fuel installed. The variations in the decay ratio and resonance frequency among five common AR methods are within 0.03 and 0.01Hz respectively. In this system, the appropriate decay ratio and resonance frequency can be estimated on the basis of Yule-Walker Method with the model order of 30.

Experiments were conducted to investigate two-phase flow instabilities in a boiling natural circulation loop with a chimney at high pressure. The SIRIUS-N facility was designed to have non-dimensional values which are nearly equal to those of a typical natural circulation BWR. The observed oscillations are found to be density wave oscillations, since the void fractions in the chimney inlet and exit are out of phase. They belong to the Type-1 category, since they occur at low flow qualities, according to the Fukuda-Kobori's classification. Moreover, the oscillation period correlates well with the passing time of bubbles in the chimney section regardless of the system pressure, the heat flux, and the inlet subcooling. Two distinct phenomena are found in relation between the oscillation period and liquid passing time in the chimney, indicating that the driving mechanisms of the instabilities are different between low and high pressures. Stability maps were obtained in reference to the inlet subcooling and the heat flux at the system pressures of 1, 2, 4, and 7.2 MPa. The flow became stable below a certain heat flux regardless of the channel inlet subcooling. The stable region enlarges with increasing system pressure. Thus, the stability margin becomes larger in a startup process of a reactor by pressurizing the reactor sufficiently before withdrawing the control rods. The obtained stability map demonstrates that the nominal operating condition of the ESBWR has a significant stability margin to the unstable region.

酸化金属を溶射したSUS304 試験片をRI 放射線源と接触させることで、酸化被膜に微弱放射線が供給され、放射線誘起表面活性による腐食緩和効果が発現した。

ABWRに9×9A型のMOX燃料を全数装荷した炉心を想定し、領域安定性試験設備SIRIUS-Fを用いて最低ポンプ速度最大出力点など広い範囲で試験を行った。その結果、同炉は運転範囲において、十分な安定性余裕があることを確認した。また許認可解析コードODYSY の減幅比(DR)および共振周波数は試験結果と精度良く一致した。

大気中においてコバルト６０によるガンマ線照射をして表面を超親水化させたジルカロイ_-_４試験体に対し、各種処理の影響を調べた。照射前の接触角が約７７度であるのに対し、線量率10 kGy/h で３４時間照射を行った場合、接触角は１０度前後まで減少し、高い親水性を示すようになる。この親水化した試験体を大気中にて24時間保持した場合には接触角の変化は認められない。さらにこの試験体を常温真空中で１時間保持したもの、および大気中100℃で１時間保持したものでは接触角が６０度前後まで上昇し、ほぼ照射前のレベルに近いところまで接触角が回復していた。一方、同じ試験体を常温純水中に保持後マグネットスターラにより１時間攪拌して流水中における影響を調べたところ、接触角は１５度前後を示し、高い親水性を維持していた。

The SIRIUS-N facility was designed and constructed for highly accurate simulation of core-wide and regional instabilities of the BWR. A real-time simulation was performed in the digital controller for heat conduction in a fuel-rod and modal point kinetics of reactor neutronics using measured void fractions in reactor core sections of the thermal-hydraulic loop. In order to estimate decay ratios, an auto-regressive method has been successfully applied for time series data of the core inlet flow rate. Experiments were conducted with the SIRIUS-N facility for the rated operating condition of 3.13 GWt natural circulation BWR. Channel and regional stability decay ratios were determined to be 0.38 and 0.54, respectively, which indicates sufficient margin for the instabilities. Experiments were extended to evaluate the stability sensitivity of design parameters such as the power profile, void reactivity coefficients, core inlet subcooling, and the fuel rod time constant.

放射線誘起表面活性（RISA）効果による防食機能を有する材料として、母材にSUS316L、中間層としてCo-CrあるいはNi-Crを用い、酸化金属皮膜ZrO2, TiO2を溶射した材料をとりあげて評価した。この材料が原子炉環境に適用できる耐久性を有するかを見極めるため、硬度評価、磨耗評価、断面評価、剥離評価等の機械的特性を実施した。また、非放射線照射下での高温高圧水環境下での耐久性も評価した。

放射線誘起表面活性（Radiation Induced Surface Activation ：RISA）効果を利用して、原子炉内部の構造材や被覆管等に同皮膜を施工することにより、(1)構造物の腐食電位を卑化させて耐食性を向上させる、(2)原子炉放射線環境下における炉心伝熱特性（限界熱流速、事故時の再冠水速度）を向上させる、ことにより経済性・安全性に優れる高性能原子炉に関する技術を開発している。ここでは、その概要について報告する。

軽水炉プラントの高経年化にともない、炉内構造材の応力腐食割れ（SCC）対策が重要課題となっている。SCC環境因子の定量的指標としてはステンレス鋼の電気化学的腐食電位(ECP)が用いられ、-230mV以下であればSCCの発生やき裂進展が抑制できるとされている。最近、室温の水溶液中において、ジルコニア(ZrO2)被覆ステンレス鋼にγ線照射することにより放射性誘起表面活性(RISA)が発現し、ECPは大きく低下することが明らかにされ、水素注入や貴金属注入に替わる新しい防食技術として期待されている。今回、BWR冷却水を模擬した高温水環境においても、RISAによるECP低下効果が顕著に発現することが明らかになったので報告する。

ABWRを模擬した領域安定性試験設備SIRIUS-Fを開発した。最低ポンプ速度最大出力点など広い範囲で試験を行った結果、許認可解析コードODYSY のDRおよび共振周波数はSIRIUS-Fの試験結果と精度良く一致した。

A new process was developed to form high performance photocatalyst film of titanium dioxide by oxidizing titanium metals in the burning flame of acetylene gases. The oxide film by this process has much superior resistance to scratch, wear and chemical reactions to a conventional titanium dioxide spray coating film. Further, this film shows superior performances in the visible-light photo-catalysis; ten times or more of photo-current density and wide absorption wave range up to 490nm comparing with the conventional spray coating film up to 410nm.

In order to investigate the conditions causing minute bubble emission boiling, critical heat flux experiments were conducted in the pool boiling system with different thermal capacities of heat transfer surfaces and method of heating.&lt;BR&gt;Stable minute bubble emission boiling was observed for a 10 mm-thick copper cylinder heated by thermal radiation. The critical heat flux obtained was 6.0MW/mm&lt;SUP&gt;2&lt;/SUP&gt;. When the heat flux exceeded above approximately 3MW/mm&lt;SUP&gt;2&lt;/SUP&gt;, a large vapor bubble formed on the heat transfer surface, then was condensed immediately and dispersed into minute bubbles.&lt;BR&gt;A 4 mm-thick silicon carbide heat transfer surface burnouted at 1.9MW/mm&lt;SUP&gt;2&lt;/SUP&gt;. When the heat transfer surface has a small thermal capacity such as the metal foil, commonly used in CHF experiments, a temporary loss of heat removal due to large bubble formation will cause a rapid temperature increase and will result in burnout.&lt;BR&gt;Minute bubble emission boiling could occur stably when the time constant determined by the heating method and thermal capacity exceeds the time required for a large bubble to condense in the subcooled fluid.

&lt;p&gt; プラズマ照射された酸化金属被膜にkGy/hのオーダーの強&lt;i&gt;γ&lt;/i&gt;線照射を行うことにより, 表面が活性化され伝熱特性の改善 (濡れ性および限界熱流束の向上) が生じる。この放射線による表面活性は放射線誘起表面活性 (Radiation Induced Surface Activation : RISA) と呼ばれ, またRISAによる沸騰熱伝達改善は放射線誘起沸騰改善 (Radiation Induced Boiling Enhancement) と呼ばれている。本稿では, 伝熱特性の向上を中心に, RISAの原理をはじめ, 腐食低減および放射線計測などの原子炉関連の応用研究の現状を解説する。&lt;/p&gt;

In order to investigate the conditions causing minute bubble emission boiling, critical heat flux experiments were conducted in the pool boiling system with different thermal capacities of heat transfer surfaces and method of heating.<BR>Stable minute bubble emission boiling was observed for a 10 mm-thick copper cylinder heated by thermal radiation. The critical heat flux obtained was 6.0MW/mm². When the heat flux exceeded above approximately 3MW/mm², a large vapor bubble formed on the heat transfer surface, then was condensed immediately and dispersed into minute bubbles.<BR>A 4 mm-thick silicon carbide heat transfer surface burnouted at 1.9MW/mm². When the heat transfer surface has a small thermal capacity such as the metal foil, commonly used in CHF experiments, a temporary loss of heat removal due to large bubble formation will cause a rapid temperature increase and will result in burnout.<BR>Minute bubble emission boiling could occur stably when the time constant determined by the heating method and thermal capacity exceeds the time required for a large bubble to condense in the subcooled fluid.

放射線誘起表面活性を発現する酸化ジルコニウムと母材との間にCoCr中間層を成膜した結果、電気的コンダクタンスが良好になり自然浸漬電位がさらに卑化し、腐食緩和効果がより高まることが判明した。

Improving the limit of boiling heat transfer or critical heat flux requires that the cooling liquid can contact the heating surface, or a high-wettability, highly hydrophilic heating surface, even if a vapor bubble layer is generated on the surface. We investigated surface wettability using metal oxides irradiated by γ-rays under normal room conditions. Contact angle, an indicator of macroscopic wettability, was measured by image-processing of the images obtained by a CCD video camera. The results showed that the surface wettability on oxidized metal pieces of titanium, zircaloy No.4,SUS-304,and copper was improved significantly by the Radiation Induced Surface Activation (RISA) phenomenon. Highly hydrophilic conditions of the test pieces were achieved after 500-kGy irradiation by ^<60>Co γ-rays. To check the RISA with radioactive metal under non-γ-ray environment use, surface wettability was investigated using metal oxides irradiated by radiation rays emitted from radioactive in the metals in room temperature conditions. The contact angle decreased linearly with integrated irradiation of neutron rays and kept in low contact angle for longer than 400 hours. These results revealed that the self-induced RISA existed with radioactive metals.[figure]

<p> プラズマ照射された酸化金属被膜にkGy/hのオーダーの強γ線照射を行うことにより, 表面が活性化され伝熱特性の改善 (濡れ性および限界熱流束の向上) が生じる。この放射線による表面活性は放射線誘起表面活性 (Radiation Induced Surface Activation : RISA) と呼ばれ, またRISAによる沸騰熱伝達改善は放射線誘起沸騰改善 (Radiation Induced Boiling Enhancement) と呼ばれている。本稿では, 伝熱特性の向上を中心に, RISAの原理をはじめ, 腐食低減および放射線計測などの原子炉関連の応用研究の現状を解説する。</p>

A gas atomization method is successfully applied for fiber and powder production from a highly viscous slag. The slag suction pressure becomes higher, when a flow guide was inserted at the exit of the gas nozzle. When the molten coal gasification slag temperature was 1500℃, a smooth continuous fiber measuring approximately 9μm in diameter was produced, which could be used as a heat insulating media and a buffer material. Powders having a diameters of approximately 9μm were produced at 1600℃, which could be used as raw material for cement.

In order to investigate the effect of considering liquid density dependence on local fluid temperature in the thermal-hydraulic stability, a linear stability analysis is performed for a boiling natural circulation loop with an adiabatic riser. Type-I and Type-II instabilities were to investigate according to Fukuda-Kobori's classification. Type-I instability is dominant when the flow quality is low, while Type-II instability is relevant at high flow quality. Type-II instability is well known as the typical density wave oscillation. Neglecting liquid density differences yields estimates of Type-II instability margins that are too small, due to both a change in system-dynamics features and in the operational point. On the other hand, neglecting liquid density differences yields estimates of Type-I stability margins that are too large, especially due to a change in the operational point. Neglecting density differences is thus non-conservative in this case. Therefore, it is highly recommended to include liquid density dependence on the fluid subcooling in the stability analysis if a flow loop with an adiabatic riser is operated under the condition of low flow quality.

Thermo-hydraulic instabilities of a boiling natural circulation loop with a chimney under high pressure were investigated using linear stability analysis. Drift-flux model was used for two-phase flow model. The instability regions as well as the thermo-hydraulic characteristics in the chimney such as wavy feature were examined, which were compared with the characteristics in low pressure. Instability could occur when exit quality was relatively low, which was the same manner as the characteristics in low pressure. In high-pressure, void was generated near channel exit, and void wave propagated in the chimney. In low pressure, steam was generated only near the chimney exit due to gravity induced flashing, and single-phase enthalpy wave, that is, temperature wave propagated in single-phase flow region. Though flow could be very stable in the high pressure and high power condition, the decay ratio of higher mode could be larger than that of lower mode.

BWR構造材の応力腐食割れ低減方策として,冷却材に水素を注入したり,構造材に貴金属を担持させることにより,腐食電位を応力腐食割れ発生のしきい値より卑化させる手法が一部試みられている。本研究ではこれらの代替として,放射線誘起表面活性効果(Radiation Induced Surface Activation)を利用して腐食電位を低下させる手法を提案する。すなわち,酸化金属皮膜にγ線などの放射線を照射することにより,軌道電子が伝導帯に励起され,同時にホールができることによりアノード電流が流れるという,非消耗型の腐食緩和手法である。

BWR構造材の応力腐食割れ低減方策として,冷却材に水素を注入したり,構造材に貴金属を担持させることにより,腐食電位を応力腐食割れ発生のしきい値より卑化させる手法が一部試みられている。本研究ではこれらの代替として,放射線誘起表面活性効果(Radiation Induced Surface Activation)を利用して腐食電位を低下させる手法を提案する。すなわち,酸化金属皮膜にγ線などの放射線を照射することにより,軌道電子が伝導帯に励起され,同時にホールができることによりアノード電流が流れるという,非消耗型の腐食緩和手法である。

A Detailed analytical model to explain the vapor film collapse was developed to evaluate the occurrence conditions of self-triggering vapor explosions. The following conclusions were drawn based on linear stability analysis using the thermo-dynamic property of water, by linearizing and perturbing basic equations (Rayleigh-Lamb-Plesset's bubble momentum equation, the mass conservation equation, the state equation for ideal gas, and the Clausius-Clapeyron equation). The vapor film stabilizes with the reduction of the hot-liquid diameter, decreasing the condensation heat transfer coefficient, and increasing the thermal radiation coefficient. The cold-liquid viscosity and surface tension have a stabilizing effect, though this effect is negligibly small where the hot-liquid diameter is over 1 mm. The analysis predicts the vapor explosion occurrence limits obtained experimentally by other researchers to within approximately 10 K.

A detailed analytical model to explain the vapor film collapse was developed to evaluate the occurrence conditions of self-triggering vapor explosions. The following conclusions were drawn based on linear stability analysis using the thermo-physical property of water, by linearizing and perturbing basic equations (Rayleigh-Lamb-Plesset's bubble momentum equation, the mass conservation equation, the state equation for ideal gas, and the Clausius-Clapeyron equation). The vapor film stabilizes with the reduction of the hot-liquid diameter, decrease of the condensation heat transfer coefficient, or increase of the thermal radiation coefficient. The cold-liquid viscosity and surface tension have a stabilizing effect, though this effect is negligibly small when the hot-liquid diameter is over 1 mm. The analysis predicts the vapor explosion occurrence limits obtained experimentally by other researchers within approximately 10 K. A simple correlation for the stability boundary is proposed by simplifying the above detailed model: the difference in cold-liquid temperature at the stability boundary between these models is less than 1 K when the condensation heat transfer coefficient is over 104W/m2·K and the hot-liquid temperature is lower than 2,000°C. © 2002 Taylor and Francis Group, LTD.

The triggering process of vapor explosion was visualized by high-speed digital video cameras at an interface between a water droplet and a molten tin pool. A cluster of bubble generated by spontaneous bubble-nucleation covered the whole contact area at 0.10ms after the impingement. Prominent fine mixing between two liquids were found to start at 0.55ms that resulting in vapor explosion. A water micro jet toward the molten tin surface might induce the fine mixing after the bubbles had expanded excessively and had contacted the subcooled water.

In order to investigate the triggering mechanism of vapor explosion, experiments were carried out in the system of a carbonate water droplet impinging onto a molten lead-bismuth pool. Carbon dioxide dissolved into water had a suppression effect on vapor explosion. The non-condensible gas reduced condensation rate, which resulted in restraining a water micro jet from impinging on to the molten lead-bismuth surface to induce the fine mixing after the bubbles had expanded excessively and had contacted the subcooled water.

Experimental studies are made on the convective heat transfer of three types of thin porous bodies. Heat transfer performances, flow patterns and temperature profiles near the porous bodies are compared with each other. The heat transfer performance of porous bodies with the largest pore diameter is large. It became clear that the high heat transfer performance depends on an excellent heat transportation ability inside the pore and near the surface of the porous bodies.

The SIRIUS facility was designed and constructed for highly accurate simulation of core-wide and regional instabilities of the BWR. A real-time simulation was performed in the digital controller for modal point kinetics of reactor neutronics and fuel-rod conduction on the basis of measured void fractions in reactor core sections of the thermal-hydraulic loop, which simulates thermal-hydraulics of a natural circulation BWR. A solid-state, series-regulated power supply, that plays a role of simulation output, was designed to attain fast response speed without loss of accuracy.<BR>A noise analysis method was performed to calculate decay ratios from dominant poles of a transfer function by applying AR method to time series of the core inlet flow rate. Experiments were conducted with the SIRIUS facility for the nominal operating condition of 3.13GWt natural circulation BWR. Channel and regional stability decay ratios were determined to be 0.38 and 0.54, respectively, which indicates sufficient margin for the instabilities. Experiments were extended to evaluate the stability sensitivity of the design parameters such as the power profile on the basis of three-dimensional steady-state analysis, the void reactivity coefficients, the core inlet subcooling, and the thermal conductance of the fuel rod.

Experiments werec conducted to investigate the triggering phenomenon in vapor explosion for various molten alloy pool depths and water droplet diameters in a droplet impingement system. The effect of the water droplet curvature Was found to be negligibly small when the droplet diameter is larger than 4.5mm. Vapor explosion conditions were identical for the depths ranging from 0.5 to 40mm. The length scale required in fragmentation and fine-scale mixing in the triggering event was considered to be less than 0.5mm.

In heat exchangers of power plants, tube supports may be exposed to scale deposition. As the explanation for this phenomenon, flashing is said to be one of the factor. In this study, flashing at the leading edge of contraction in a rectangular flow path were observed. Pressure drop and steam void characteristics at the flashing occurrence point were measured. The phenomenon were also investigated by photographs. Between two types of contraction, with same contraction ratio and different hydraulic diameter, considerable difference of boundary temperature of flashing occurrence were seen. Most of the bubbles detected at the leading edge of contraction were very small while those at the upstream included bubbles of many sizes.

A Noise analysis method was performed to calculate decay ratios from dominant poles of a transfer function by applying AR method to time series of the core inlet flow rate. Experiments were conducted with the SIRIUS facility, which simulates a representative natural circulation BWR. Channel and regional stability decay ratios at the nominal operating condition were determined to be 0.38 and 0.54, respectively, which indicates sufficient margin for the instabilities. Experiments were extended to investigate the effects of the design parameters on stability.

Sustainable small-scale spontaneous vapor explosion was proposed to utilize for the innovative rapid cooling and the liquid atomizing processes, namely the CANOPUS method. On the basis of CANOPUS method, a highly viscous coal gasification slag was atomized into 30μm powders for use of cement raw materials. The cooling rate of Al_<89>-Si_<11> particle was up to 1.5×l0^8K/s, which is more than 280 times higher than the conventional cooling method.

The triggering process of vapor explosion was visualized by high-speed digital video cameras at an interface between a water droplet and a molten alloy pool. A ring of 3mm diameter was seen at 0.01ms after the impingement and is considered to be a cluster of bubbles generated by spontaneous bubble-nucleation. The cluster of bubble covered the whole contact area at 0.05ms. Prominent fine mixing between two liquids were found to start at 0.70ms that resulting in vapor explosion.

Thermo-hydraulic instabilities of a boiling natural circulation loop with a chimney under low and high pressure were investigated using linear stability analysis. The effect of nuclear coupling was also considered. Both in low- and high-pressure conditions, instability could occur when exit quality was relatively low. In low-pressure condition, flashing only near the exit of the chimney could induce instability, and enthalpy wave of single-phase flow was generated in the chimney. In high-pressure condition, void was generated near channel exit, and void wave propagated in the chimney. In the high pressure and high power condition, though flow could be very stable, the decay ratio of higher mode could be larger than that of lower mode. The sensitivity on decay ratio to the thermal power, inlet subcooling, void reactivity feedback coefficient and so on could be very low when there was a long chimney.

Liquid metal flows were visualizaed by using neutron radiography. Single-phase lead-bismuth eutectic flow was visualized by a tracer method. High speed visualization of water evaporation behaviors in lead-bismuth-tin alloy was carried out. Thermal neutron radiography system at JRR-3M in JAERI was used. It was shown that neutron radiography was applicable to visualize single-and two-phase flow of heavy metal.

Liquid metal flows were visualizaed by using neutron radiography. Single-phase lead-bismuth eutectic flow was visualized by a tracer method. High speed visualization of water evaporation behaviors in lead-bismuth-tin alloy was carried out. Thermal neutron radiography system at JRR-3M in JAERI was used. It was shown that neutron radiography was applicable to visualize single-and two-phase flow of heavy metal.

Experiments have been conducted to investigate thermal-hydraulic instabilities at system pressure ranging from 1 to 7.2 MPa in a boiling natural circulation loop with a chimney. Stability maps in reference to the system pressure, the channel inlet subcooling, and heat flux are presented. Two different types of instabilities were observed in the facility at relatively low and high system pressures. Both of the instability mechanisms were clarified by investigation of the transient flow pattern and the response of the driving force of the circulation to momentum energy.

The diverter surface of the ITER Fusion Engineering Reactor is exposed to strong radiation locally up to 20 MW/m&lt;SUP&gt;2&lt;/SUP&gt;. We have proposed a diverter cooling system which consists of concave surfaces cooled by two-dimensional liquid jets. Experiments are conducted to investigate local heat transfer coefficeint and critical heat flux on flat and concave surfaces under various cooling conditions. Based on photographic study, a critical heat flux model was derived by modifying a Haramura&#039;s model to take account of the subcooling effect. The proposed correlation based on this model was in good agreement with the experimental data.

Instability of a boiling natural circulation loop with a chimney at low pressure and low heater power was investigated by linear stability analysis and experiment. A homogeneous and thermodynamic equilibrium model for two-phase flow was used. The effect of flashing induced by pressure drop in the heated channels and the chimney was considered. The effects of coupling between two boiling channels were investigated. It was found that in-phase-mode instability was apt to occur when channel inlet subcooling was large and boiling began in the chimney. In-phase-mode instability easily occurred when channel length became short and the chimney became long. Out-of-phase-mode instability was apt to occur when chimney length became small and boiling began in the channel. It was suggested that in-phase-mode instability was density weve oscillation induced by flashing in the chimney and out-of-phase-mode instability was density wave oscillation induced by boiling in the channels. The analytical results agreed qualitatively with experimental results.

Experiments are are conducted to investigate two-phase flow instabilities in a boiling natural circulation loop with a chimney induced by flashing in the chimney at lower pressure. The type of instability that occurred in the experiments is suggested to be density wave oscillations induced by flashing in the chimney. The differences from other instabilities such as geysering, flow pattern transition instability, and natural circulation oscillations are discussed on the basis of the dynamic characteristics, the oscillation period, and the transient flow resume.

As a candidate for an innovative steam generator for fast breeder reactors, a heat exchanger with direct contact heat transfer between melting alloy and water was proposed. The evaluation of heat transfer characteristics of this heat exchanger is one of the research subjects for the design and development of the steam generator. In this study, the effect of the pressure on heat transfer characteristics and the required degree of superheating of melting alloy above water saturation temperature are evaluated during the direct contact heat transfer experiment by injecting water into Wood's alloy. In the experiment, the pressure, the temperature of the Wood's alloy, the flow rate of feed water, and the depth of the feed water injection point are varied as parameters. As a result of the experiment, the product of the degree of Wood's alloy superheating above water saturation temperature and the depth of the feed water injection point is constant for each pressure. This constant increases as the pressure rises.

Instability of a boiling natural circulation loop with a chimney at low pressure and low heater power was investigated by linear stability analysis and experiment. A homogeneous and thermodynamic equilibrium model for two-phase flow was used. The effect of flashing induced by pressure drop in the heated channels and the chimney was considered. The effects of coupling between two boiling channels were investigated. It was found that in-phase-mode instability was apt to occur when channel inlet subcooling was large and boiling began in the chimney. In-phase-mode instability easily occurred when channel length became short and the chimney became long. Out-of-phase-mode instability was apt to occur when chimney length became small and boiling began in the channel. It was suggested that in-phase-mode instability was density weve oscillation induced by flashing in the chimney and out-of-phase-mode instability was density wave oscillation induced by boiling in the channels. The analytical results agreed qualitatively with experimental results.

Experiments are are conducted to investigate two-phase flow instabilities in a boiling natural circulation loop with a chimney induced by flashing in the chimney at lower pressure. The type of instability that occurred in the experiments is suggested to be density wave oscillations induced by flashing in the chimney. The differences from other instabilities such as geysering, flow pattern transition instability, and natural circulation oscillations are discussed on the basis of the dynamic characteristics, the oscillation period, and the transient flow resume.

The diverter surface of the ITER Fusion Engineering Reactor is exposed to strong radiation locally up to 20 MW/m². We have proposed a diverter cooling system which consists of concave surfaces cooled by two-dimensional liquid jets. Experiments are conducted to investigate local heat transfer coefficeint and critical heat flux on flat and concave surfaces under various cooling conditions. Based on photographic study, a critical heat flux model was derived by modifying a Haramura's model to take account of the subcooling effect. The proposed correlation based on this model was in good agreement with the experimental data.

The large eddy simulation (LES) method is applied to turbulent natural convection in a horizontal concentric annulus. Numerical results are obtained for Rayleigh numbers based on the gap width from 2.51&amp;times;10&lt;SUP&gt;6&lt;/SUP&gt; to 1.18&amp;times;10&lt;SUP&gt;9&lt;/SUP&gt;. These results are compared with other researchers&#039; experimental data and the validity of LES is investigated. It is found that LES is very useful even if at high Ra, at which the DNS is not applicable. However, in the cases where temperature differences between cylinders are large, effect of dependency of physical properties on temperature is not negligible, which causes the poor coincidence betwenn the experiments and the numerical results.

The large eddy simulation (LES) method is applied to turbulent natural convection in a horizontal concentric annulus. Numerical results are obtained for Rayleigh numbers based on the gap width from 2.51&times;10⁶ to 1.18&times;10⁹. These results are compared with other researchers' experimental data and the validity of LES is investigated. It is found that LES is very useful even if at high Ra, at which the DNS is not applicable. However, in the cases where temperature differences between cylinders are large, effect of dependency of physical properties on temperature is not negligible, which causes the poor coincidence betwenn the experiments and the numerical results.